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DOE OSTI2016

Development of Ultra Small Scale Mechanical Testing and Localized He Implantation for Nuclear Applications

Mechanical properties degradation of components deployed in nuclear applications has been a concern since the deployment of first generation nuclear power plants. Radiation damage along with environmental degradation, elevated temperatures, and long term deployment of nuclear structures pose a challenging environment for all materials in these applications. Traditionally, the mechanical properties are assessed by standardized macroscopic testing methods. While those are providing engineering relevant data these techniques do have limitations, especially on reactor irradiated materials, due to the large size of the samples. Limitations on the amount of radioactivity of large samples must be applied for research applications as they are conducted at universities. In addition it can be difficult to retrieve sufficiently large amounts of material from an operating reactor component. Therefore smaller sample sizes are desirable and reduce the amount of radioactive material which needs to be handled or material which needs to be retrieved. In addition, the number of tests which can be performed on materials irradiated in a test reactor program is limited due to limited space in specific dose and temperature positions. Downsizing of the mechanical test can overcome this issue and perform more tests on the same amount of irradiated materials, reducing the statistical uncertainties. In addition, smaller scale samples allow access to specific regions of interest in a material or component, be it the zone in front of a crack tip or the sampling of very small components like springs or welded areas. In addition ion beam irradiation has become an integral part of nuclear materials research to enhance the understanding of radiation damage in materials in a time and cost effective fashion. The drawback of ion beam irradiation is the small penetration depth of the ions into the material making localized microstructural characterization as well as small scale mechanical testing necessary. Another more fundamental motivation is the fact that since the beginning of electron microscopy, one has been concerned about imaging and quantifying defects in materials. Multiscale tools to image a large range of radiation induced defects have been established starting from atom probe tomography to transmission electron microscopy, up to scattering techniques. Mechanical testing, however, has always been performed on a large scale, and an attempt to match the same scale where a defect is observed with the materials' mechanical property response has not yet been realized. However, the question of separate effects testing on specific radiation induced defects remains and is an additional driving force for small scale mechanical testing. In this work we present an approach for small scale mechanical testing in irradiated materials and the variety of possible test methods developed today. We also present recent efforts of performing localized ion beam implantation to achieve a number of different doses in the same region of interest enhancing comparability and quantification of observed phenomena. It is the intention to give the reader an overview of the different microscale mechanical testing techniques possible today as well as an introduction of a novel ion beam irradiation method using a newly established capability. In this paper we can't discuss all aspects of all small scale mechanical testing of irradiated materials However, It does give an overview over the techniques enabled due to recent effort and enhanced instrument capability. While indentation or micro compression testing has been performed previously tensile testing and ion beam irradiation using the He ion beam microscope are novel approaches not attempted previously. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2021

Use of nucleic acid amplification testing for rapid detection of Mycobacterium tuberculosis complex among U.S. tuberculosis patients from 2011–2017

Background: Nucleic acid amplification (NAA) tests rapidly detect Mycobacterium tuberculosis complex directly from clinical specimens, providing valuable results for those evaluated for tuberculosis. Methods: We analyzed characteristics of cases with NAA testing performed, compared cases with positive and negative NAA test results, and calculated turnaround time and time to treatment for all verified cases reported to the National Tuberculosis Surveillance System in the United States during 2011–2017. Results: Among 67082 verified tuberculosis cases with NAA testing information, 30820 (45.9%) were reported as not having an NAA test performed; the proportion without NAA testing declined annually, from 60.5% in 2011 to 33.6% in 2017. Of 67082 verified cases, 27912 (41.6%) had positive, 8215 (12.2%) had negative, and 135 (0.2%) had indeterminate NAA test results. Among the 33937 cases with an acid-fast bacilli (AFB) smear-positive result, 24093 (70.9%) had an NAA test performed; 11490 of the 30244 (38.0%) with an AFB smear-negative result had an NAA test performed. Although sputum was the most common specimen type tested, 79.8% (7023/8804) of nonsputum specimen types had a positive NAA test result. Overall, 63.7% of cases with laboratory testing had NAA test results reported <6 days following specimen collection; for 13891 cases not yet on treatment, median time to treatment after the laboratory report date was 2 days. Conclusions: Our analyses demonstrate increased NAA test utilization between 2011 and 2017. However, a large proportion of cases did not have an NAA test performed, reflecting challenges in broader uptake, suggesting an opportunity to expand use of this diagnostic methodology.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2016

MO-AB-206-00: Nuclear Medicine Physics and Testing

This education session will cover the physics and operation principles of gamma cameras and PET scanners. The first talk will focus on PET imaging. An overview of the principles of PET imaging will be provided, including positron decay physics, and the transition from 2D to 3D imaging. More recent advances in hardware and software will be discussed, such as time-of-flight imaging, and improvements in reconstruction algorithms that provide for options such as depth-of-interaction corrections. Quantitative applications of PET will be discussed, as well as the requirements for doing accurate quantitation. Relevant performance tests will also be described. Learning Objectives: Be able to describe basic physics principles of PET and operation of PET scanners. Learn about recent advances in PET scanner hardware technology. Be able to describe advances in reconstruction techniques and improvements Be able to list relevant performance tests. The second talk will focus on gamma cameras. The Nuclear Medicine subcommittee has charged a task group (TG177) to develop a report on the current state of physics testing of gamma cameras, SPECT, and SPECT/CT systems. The report makes recommendations for performance tests to be done for routine quality assurance, annual physics testing, and acceptance tests, and identifies those needed satisfy the ACR accreditation program and The Joint Commission imaging standards. The report is also intended to be used as a manual with detailed instructions on how to perform tests under widely varying conditions. Learning Objectives: At the end of the presentation members of the audience will: Be familiar with the tests recommended for routine quality assurance, annual physics testing, and acceptance tests of gamma cameras for planar imaging. Be familiar with the tests recommended for routine quality assurance, annual physics testing, and acceptance tests of SPECT systems. Be familiar with the tests of a SPECT/CT system that include the CT images for SPECT reconstructions. Become knowledgeable of items to be included in annual acceptance testing reports including CT dosimetry and PACS monitor measurements. T. Turkington, GE Healthcare.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

MO-AB-206-02: Testing Gamma Cameras Based On TG177 WG Report

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Correlation of Qualification and Accelerated Testing with Field Degradation

Qualification testing of solar photovoltaic (PV) modules per IEC 61215 standard encompasses a set of well-defined accelerated stress tests (irradiation, environmental, mechanical and electrical) with strict pass/fail criteria based on functionality/performance, electrical and mechanical safety, and visual requirements. The qualification testing does not identify all the possible reliability failures or durability issues witnessed in the actual field. However, it does identify the major/catastrophic design quality issues (infant mortality), which could occur early in a module’s field life. Unfortunately, in the industry, there is no consensus-based standard or protocol available to validate manufacturers’ warranty period claims based on the end-of-life wear-out failure mechanisms. Being able to predict the field degradation rates and lifetime using mode-specific accelerated tests on moderately field-aged modules, for example 10 years, for certain commonly occurring degradation modes would be a powerful contribution to all the stakeholders of the PV industry. There is a need to formulate guidelines to find the correlation of accelerated testing with field degradation for individual degradation modes observed in the field. Potential approaches to fulfill this need were experimented in this project and are presented in this report. Two of the most commonly seen degradation modes – encapsulant browning and solder bond thermomechanical fatigue – are addressed in this report through presenting the accelerated testing methods and physical modelling approaches. Two accelerated test methods (UV and thermal cycling) were designed to induce the wear-out mechanisms in the modules and not infant mortality failures. To focus on the wear-out failure modes, the tests were performed on field-aged modules, instead of fresh modules, with known degradation rates and degradation modes. The degradation rates were measured based on the degradation rates of individual cell parameters (short-circuit current, I sc and series resistance, R s ) instead of the degradation rate of maximum power (P max ) which cannot be used to study any specific/limited degradation mode as the P max degradation could be caused by many different degradation modes. The key characteristic parameter considered for encapsulant discoloration and solder bond degradation is I sc and R s , respectively. The developed physical models/equations relating the testing degradation data and field weather data were utilized to determine the acceleration factor (AF) and to predict the degradation rate for a specific degradation mode of PV modules installed in two different climates, Arizona (hot) and New York (cold). The predicted degradation rates for encapsulant browning and solder bond degradation have also been validated against the actual measured field degradation rates. UV testing cost reduction: To determine acceleration factor, single UV chamber was used for accelerated UV testing at three different temperatures simultaneously, instead of three chambers for three temperatures (or one chamber sequentially which is an extremely time-consuming approach). A passive heating approach was successfully used to achieve and maintain multiple temperatures in the accelerated UV chamber. Foam insulation boards of varying thicknesses were utilized to obtain three different temperatures for UV testing. However, these foam materials tend to disintegrate in the presence of moisture. To alleviate this problem, silicone heating blankets with temperature controls and humidifier can be used if the UV testing is to be performed under hot-humid conditions instead of conventional hot-dry conditions. Thermal cycling testing cost reduction: Single thermal cycling chamber was used for accelerated thermal cycling testing at three different upper temperatures simultaneously, instead of three chambers for three upper temperatures (or one chamber sequentially which is an extremely time-consuming approach). An active heating approach based on silicone heating blankets was successfully used to achieve and maintain multiple upper temperatures in the accelerated thermal cycling chamber. Such blankets provide stable temperature control and thereby more reliable and repeatable for replicate experiments.

14 SOLAR ENERGY↗

DOE OSTI2017

Aspects of Assessment of Packages with Wood Filled Impact Limiters during Fire Tests - 17191

Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by steel sheets. These impact limiters shall ensure that the transport casks meet the mechanical and thermal IAEA regulatory test requirements. According to the accident conditions of transport it is mandatory to expose the specimens to a cumulative effect by mechanical and thermal impacts. The mechanical tests consist of a free drop from 9 m onto a flat unyielding target and a 1 m drop onto a puncture bar. After damage caused by mechanical test sequences the package has to withstand a severe fire scenario. Corresponding to the IAEA advisory material it is required that the impact attitudes for the 9 m drop test and for the puncture test have to be such as to produce maximum damage, taking into account the thermal test. Moreover, any damage, which would give rise to increased radiation or loss of containment or affect the confinement system after the thermal test, should be considered. During and following the thermal test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Different works from the French Institute for Radiological Protection and Nuclear Safety (IRSN) and BAM show that additional energy supply from a pre-damaged impact limiter to the cask could occur. This effect should be considered within the safety assessment of the containment. Thermal effects at the closure system of the cask, which might result in an elevated activity release, have to be excluded. BAM conducted small scale tests with wood filled metal buckets showing continuing combustion processes during the cooling down phase. These test results are presented. As not much is known about smouldering processes in wood filled impact limiters, it is highly complex to define pre-damage of impact limiters, which are conservative, regarding the maximum damaging energy flow from the impact limiter to the containment system. More research has to be done to develop models to examine the effects of smouldering impact limiters on the containment of packages for the transport of radioactive material. Aspects of assessment and its difficulties are shown. BAM as a competent authority for the approval of transport casks for radioactive material in Germany operates the test facilities to examine the issue of mechanical damage, combustion and heat transfer for such kind of package systems. For this purpose, the knowledge from real drop tests with casks of a mass partly over 100 tons was transferred to a test application. A thermal test will take place with a wood filled test specimen with a diameter of about 2.3 meters. The aim is to understand the phenomena of smouldering under the consideration of relevant regulatory boundary conditions. The process of smouldering is described with regard to the requirements in the thermal assessment of safety of packages for the transport of radioactive material. Requirements concerning the pre-damage of packages for the maximum damage of impact limiters are discussed. Parameters influencing the smouldering process are identified. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Validation Testing for Molten Chloride Reactor Experiment Equipment Removal and Disposal Techniques

The Molten Chloride Reactor Experiment (MCRE) will be the first reactor featuring a fast-spectrum molten chloride circulating nuclear fuel system in the world. Planning for equipment removal and disposal (ERD) of MCRE has identified several technology gaps due to the unique environment of this nuclear experiment. Some of the gaps arise from the application of existing disassembly and/or sizing methods to novel material forms or in novel configurations. Others arise from unknown material behavior. This paper summarizes proposed test plans for ERD validation experiments to address these complicated or unknown equipment removal procedures. At the Waste Management Symposia in 2024, the Idaho National Laboratory (INL) MCRE ERD team presented the challenges associated with hosting multiple nuclear experiments in series with only brief transition periods between systems. Such difficulties include higher dose rates, the presence of radioisotopes infrequently encountered in reactor decommissioning and radioactive waste management, lack of intrinsic remote-operations infrastructure in the test bed, space constraints in the test bed, and contamination minimization requirements. To address these challenges, remote or semi-remote technologies are planned to be implemented in a non-hot cell environment with limited space availability. The team also discussed how a systems engineering approach is being used for conceptual development and design of equipment removal systems to address these challenges. For example, to reduce constraints for the removal of more difficult components, non-activated, noncontaminated elements are planned to be taken out first where possible. Still, there are complexities associated with the remaining components. In this work, the operational framework for MCRE ERD was reviewed for technical gaps and open questions, and test plans were drafted to address these areas. The tests plans were written for the following categories: vision systems, pipe cutting, drill/grout/filler, flush salt, and miscellaneous, with the miscellaneous group consisting of tests like techniques for removing bearings and reflector bricks. The test plans explore material, infrastructure, and staffing requirements needed for test execution. The test plans additionally focus on the evaluation of success. Determining the outcome of a test is imperative - as these explorative actions have the potential to rearrange or re-scope planned ERD activities. Success criteria identified thus far include required tool output, required area(s), debris production and mitigation, and repeatability. Test plans are an essential aspect of the systems engineering approach to MCRE ERD. They are used as the beginning steps in defining use cases for the ERD system. Performance of the validation tests is expected to begin in the summer of 2025 and will take approximately 9 to 12 months to complete. Execution of these plans will be expedited by specifying test needs ahead of time, facilitating efficient interactions with any subcontractors tasked with running the requested tests. Evaluating the outcomes of these tests will inform MCRE ERD procedures and timing and will also identify additional technical constraints for the MCRE ERD System. This upfront process optimization effort will help the project save time and resources at the end of the experiment.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2015

CIRFT Testing of Used Nuclear Fuel Rods from the Limerick Generating Station Boiling Water Reactor (FY15 Status Report)

The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 10 7 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m -1 . The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles-to-failure ranged from 1.22 × 10 5 to 4.70 × 10 6 , and the amplitudes varied from 15.2 to 7.6 N·m. The measurements at the interrupts indicated a range of flexural rigidity from 30 to 50 N·m 2 . Online monitoring revealed that the flexural rigidity was slightly lower due to the higher loading, from 25 to 42 N·m 2 . Generally, no substantial change of rigidity was observed based on the online monitoring during the cyclic fatigue testing process. Overall, the decreasing trend of sample lifetime with increasing amplitude is well defined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Design and Testing for a New Thermosyphon Irradiation Vehicle

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heat loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

RCCS Testing with the Water-based NSTF: Year-1 Single-Phase Results

Under support from the Department of Energy (DOE) and the Office of Advanced Reactor Technologies (ART), a large scale test facility has been constructed to generate validation data for passive decay heat removal systems of advanced reactors. Located at Argonne National Laboratory, the Natural convection Shutdown heat removal Test Facility (NSTF) reflects key features of a ½ scale, water-based, Reactor Cavity Cooling System (RCCS). This report serves as a summary of the single-phase testing results obtained during the programs first year of matrix testing. An introduction to the high-level program objectives is first presented, followed by a brief summary of facility design, instrumentation, and control systems. Details of the NQA-1 driven test method are included, which govern the programs quality assurance requirements and provide a background for the testing strategy and procedures. Checkout activities, including verification of gross system mass and energy balances, heat loss quantification, and isothermal facility characterization, were performed early in the first year and results presented in this report. These tests have identified relevant facility characteristics such as K-factors, frictional pressure losses, parasitic heat loss rates, and transient heating response. Results are then presented from four accepted matrix tests where the facility established a steady-state, single-phase mode of flow operation. All matrix tests were performed with an identical top-level procedure: first initiating a heater power ramp to target levels, and then holding power levels constant until the facilities structural members and network piping reached thermal equilibrium. Upon obtaining stable thermal and hydraulic conditions, active cooling was initiated off of the bulk storage tank to maintain constant riser inlet temperatures of 30°C. This steady-state period was then held for a minimum required duration, at which point the heaters were ramped down and test concluded. The two baseline tests completed reflect conditions prototypic to the full scale plant at both normal operation, and design basis Depressurized Loss of Forced Cooling (DLOFC) conditions. These parameters are defined by full RCCS decay heat loads of 1.4 MWt and 2.1 MWt, which translate to NSTF scale thermal power levels of 34.4 kWt and 51.6 kWt, respectively. Two additional tests were also conducted, which repeated baseline cases but included modification by single parameter variations in decay heat load, or heater power, at 700 kWt, and 2.8 MWt. Finally, a scoping test was performed to demonstrate the facilities ability to operate in a two-phase mode of operation. The facility successfully operated for a 3-hour period at this mode, reaching peak inventory temperatures of 105°C and condensing approximately 50 gallons of boil-off inventory. The testing parameters for this demonstration were based off the design basis DLOFC condition, and will be the reference for new baseline cases as the program moves forward with two-phase matrix testing.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Federal Emergency Management Agency Mobile Housing Unit Running Gear Testing (Final Report)

The Federal Emergency Management Agency (FEMA) stages manufactured housing units (MHUs) at key locations to be used during disasters when other housing accommodations cannot be arranged for individuals or families. Due to the unpredictability of natural disasters and the limitations of the manufactured housing industry to produce large numbers of MHUs in very short periods of time, it is necessary for FEMA to store a large number of units for many months or years. With these long storage periods come degradation issues with frames, tires, bearing, and braking systems. These types of issues are also of interest to the U.S. Department of Transportation (DOT) Federal Motor Carrier Safety Administration (FMCSA) because regulated carriers normally transport these units and 49 CFR 393.75 Tires specifies tire loading restrictions for manufactured homes. To compare the running gear (tires, brakes, drums, bearings, axles, springs, and spring mounting hardware) currently specified in the 2014 Rugged Based Performance Requirements (RBPR) and the newly recommended running gear specified in the upcoming revision to the RBPR, the Oak Ridge National Laboratory (ORNL) outlined a Longevity Test Plan, which focused on the assessment of component longevity (failure rates by component type). The Test Plan also studied the forces imparted to the MHU from the running gear from interaction with the roadway for both the running gear in the 2014 RBPR and the running gear specified in the upcoming RBPR revision. Since the assessment of component longevity and failures are also issues of concern for FMCSA, both agencies collaborated to support the research presented in this report. The tests and data collection were conducted at the Navistar Proving Grounds (NPG) in New Carlisle (near South Bend), Indiana, during the May-November 2018 period. Four alternatives for the suspension system of a typical three-bedroom MHU were considered in this test. Alternative 0 was the MHU that was selected from the FEMA staging area in Selma, Alabama. Empirical data had shown that, on average, this alternative experiences one tire failure every 150 miles (or 0.0067 tire failures per mile traveled). Before testing, Alternative 0 was equipped with similar running gear hardware, although in new condition and installed following the manufacturer’s specifications. This became the Baseline Unit, or Alternative 1. Another unit, also selected from the FEMA staging area in Selma, Alabama, was equipped with better axles (i.e., axles that were rated for 7,000lbs as opposed to 6,000lbs for the Baseline Unit) and better tires. This became the Upgrade Unit, or Alternative 2. At the end of the Longevity Test, the Upgrade Unit was mounted with radial tires. This became Alternative 3.Alternative 1 (Baseline Unit) had, on average, one tire failure every 251 miles (or 0.0040 tire failures per mile traveled). So, by simply following the correct manufacturer’s specifications for the assembly of the transportation system, a reduction of 41% in the per mile tire failure was achieved. In the same test, Alternative 2 (Upgrade Unit) showed, on average, one tire failure every 728 miles traveled (or 0.0014 tire failure s per mile traveled). When the Upgrade Unit was mounted with radial tires in the last part of the test, no tire failures were observed for the 2,400 miles tested. That is, in the case of the radial tires (Alternative 3) the tire failure was less than one per 2,400 miles traveled (or less than 0.0004 tire failures per mile traveled). The data collected for the Longevity Test showed that the maximum accelerations were registered in all cases at the center of the MHU in the vertical direction. The overall maximum was registered by the Baseline Unit, Alternative 1. If that value is normalized to 100%, then the Upgrade Unit (Alternative 2) showed, on average, 81% at the same location, and Alternative 3 (radial-tire test) 59%. That is, the upgraded transportation system showed a reduction of 19% and 41% in the vertical accelerations transmitted to MHU at its center when regular and radial tires, respectively, were used. As part of the Test Protocol, NPG conducted a visual inspection of the MHUs and transportation system at the end of each test day. ORNL reviewed all the inspection reports in an attempt to determine if there was any difference in the damage that the Baseline and Upgrade MHUs showed during the test. The Upgrade Unit always showed lower accelerations transmitted to the MHU than the Baseline Unit. In most cases, these differences were statistically significant. The FEMA inspection report, conducted by a FEMA certified inspector at the end of the tests, included a shorter list of damaged items for the Baseline Unit than for the Upgrade Unit, although similar type of damages was described for both units. This may be an indication that, although lower than those of the Baseline Unit, the accelerations transmitted by the Upgrade Unit were above the threshold at which MHU elements start to fail.

42 ENGINEERING↗

DOE OSTI2020

Analysis of Air-Purifying Respirator (APR) and Powered Air-Purifying Respirator (PAPR) Cartridge Performance Testing on a Hanford AP Tank Farm Exhauster Slipstream (Volume 1)

Washington River Protection Solutions (WRPS) tested four types of chemical cartridges for use in air-purifying respirators (APR) and powered air-purifying respirators (PAPR). These tests were undertaken to determine the period of time that the cartridges would provide adequate performance1 for APRs and PAPRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the Hanford AP tank farm exhauster slipstream. The Occupational Safety and Health Administration (OSHA) considers cartridge testing to be a valid approach for establishing cartridge change schedules. Testing commonly is 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. Cartridge testing on a slipstream from the Hanford AP tank exhauster was conducted on March 23-24, 2018. This testing focused on both APR and PAPR cartridges. Previous testing of APR cartridges was conducted on the AP exhauster in June of 2016. However, an AP exhauster upgrade was completed in September 2016. In the most recent testing, slipstream vapors from the new AP exhauster were fed to two respirator cartridge test stands, one for the PAPR respirator cartridges and the other for the APR respirator cartridges. Both the APR and PAPR test stands were developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Multipurpose APR cartridges—SCOTT 7422-SD1 and SCOTT 7422-SC1 (SCOTT Safety, Monroe, North Carolina)—were assessed on separate days using the APR cartridge test stand. Multipurpose PAPR cartridges—MSA OptiFilter TL (MSA Safety Inc., Pittsburgh, Pennsylvania) and 3M FR-57 (3M Company, Maplewood, Minnesota)—were also tested consecutively over the same two days as the APR cartridge tests, using the PAPR cartridge test stand. Sample media (i.e., sorbent tubes) were used to collect samples of the vapor stream entering and exiting the respirator cartridges, and the samples then 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. The key conclusions from the analysis are described below.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2023

FY23 Report on Water NSTF Testing at Two-Phase Conditions: Off-normal Scenarios

Under support from the Department of Energy (DOE) and the Office of Advanced Reactor Technologies (ART), a large-scale test facility has been constructed at Argonne National Laboratory to generate NQA-1 qualified validation data for passive decay heat removal systems in advanced reactors. The Natural convection Shutdown heat removal Test Facility (NSTF) reflects key features of a ½ scale, water-based, Reactor Cavity Cooling System (RCCS) and is intended to study the behavior, bound performance, and ultimately guide design decisions for passive decay heat removal systems for advanced reactors. In addition to the experimental activities detailed in this report, a supportive computational modeling effort is on-going which has been demonstrated to significantly strengthen the experimental program while also improving accuracy of the computer models. Together these create a mutually beneficial relationship integral to meeting the overall program objective of examining the heat removal performance of the RCCS concept. This report details the experimental activities and upgrades performed during the program’s fifth year of water-based operation. The theme of this year of testing was examining off-normal scenarios. In practice, that meant examining scenarios with blockages or other design basis conditions. From a maintenance and capability standpoint that meant assessing valving options for implementing blockages, assessing drain and refill capabilities, and ensuring in all cases that operators could safely, efficiently, and repeatably bring the facility to an off-normal condition and return it to the normal operating state. In one notable case, no solution existed, nor was a solution commercially available, that met the facility needs for implementing blockages within the two-phase region of the chimney. A valve and feedthrough were custom-designed for NSTF to operate submerged in saturated water while being safely actuated from outside the tank. The fifth year of testing included eight matrix tests consisting of 162 hours of active heating, 9,735 kWh of electrical heating, with seven tests classified as Accepted per NQA-1 and one as Trending. All matrix tests were performed at two-phase flow conditions. The power parametric series from previous years was extended to include another scenario with a decay load equivalent to 1.75 MWt, full scale, further resolving the system flow oscillation response as a function of input power. A depletion test was performed to extend previous work and examine the system’s approach to stagnation and geysering that were not previously observed. This test began at an initial inventory level of 50%, prototypic power of 2.1 MWt, and employed an accelerated drain of 0.75 gpm plus boiloff until stagnation conditions were achieved and multiple geysering events were recorded. Further exploring the geysering phenomenon was a separate effects test examining static boiling in the risers and chimney. This static boiling test was performed at 48 kWe, a higher power than testing in the previous year, and included enhancements to the drain system to remove inventory in a better-controlled and more reliable manner. Two tests were performed, at prototypic conditions of 1.4 MWt and 2.4 MWt, respectively, examining the system response to throttled conditions in the singlephase region of the piping. A ball valve immediately before he riser inlet header was increasingly throttled, progressing the system through a repeatable pattern of two-phase oscillations and stable regimes before ultimately inducing stagnation and geysering. Concluding the year, two tests were performed at baseline conditions of 70% initial inventory level and 2.1 MWt of prototypic decay heat to examine both repeatability and the response to throttling within the two-phase region, at the inventory tank inlet using the custom valve mentioned above. The first such test repeated transient operating conditions and revealed significant sensitivity t

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Radiometric Testing of Germicidal UV Products, Round 1: UV-C Towers and Whole-Room Luminaires (CALiPER Summary Report)

This summary report analyzes the independently tested performance of 13 germicidal ultraviolet (GUV) products purchased between February and July 2022. A companion full report (DOE 2023) provides additional information and discussion of the tested products, test methods, and results. The products were of three different types: • Seven portable, consumer-oriented GUV towers designed to be placed on the floor or a desk of an unoccupied room to disinfect air and surfaces. Five of these products used LED sources and two products had low-pressure mercury (LPM) sources. • One GUV whole-room luminaire designed to be installed on a ceiling to disinfect air when a room is occupied. This product had LED sources. • Five GUV troffer or high-bay style whole-room luminaires designed to be installed in or suspended from a ceiling to disinfect air and surfaces when a room is unoccupied. All five had LPM sources. Product testing covered radiometric and electrical performance for all 13 products as well as photobiological safety evaluation if product documentation included testable claims. Measurement results enable comparison between products and against manufacturer or vendor claims. Testing identified numerous issues related to the accuracy of claimed GUV product performance. Claims were often untestable, contradictory, ambiguous, or used incorrect units and/or terminology. When claims were testable, they often did not match test results. For example, three LED products that claimed to emit UV-C emitted only UV-A. Product claim issues were more common among consumer-oriented tower products, but all product types exhibited problems with accurate performance claims. The UV-C radiant efficiency (calculated as UV-C output power divided by electrical input power) of the products varied widely, even among similar products using the same source technologies. For example, the UV-C radiant efficiency of LPM products varied by greater than a factor of three for the same product type, indicating a large potential energy savings opportunity for products that are better designed for efficiency. LED products had orders-of-magnitude lower UV-C radiant efficiency than LPM products. This study also identified several testing challenges and limitations. Most significant among these is the capability to accurately test and report the performance of larger GUV products. Whereas integrating spheres are used to quickly measure total radiant flux (i.e., output power) and spectral distribution, goniometers are used to measure radiant intensity distribution (from which radiant flux can be calculated). Integrating spheres require a specialized and costly coating to test UV, and the testing laboratory for this round of products had only a 20-inch diameter hemisphere with this capability. The integrating sphere accommodated just 2 of the 10 UV-C emitting products. Goniometer testing had a different size limitation in that mirrors typically used to increase goniometer test distance to the far field reflect little to no UV. As a result, the study evaluated only 6 of 13 products in the far field. Electronic files of UV-C intensity data for the other 7 products, which would typically be imported into design software for designing GUV applications, may not be reliable for predicting irradiance at arbitrary far-field distances (IES 2022a; CIE 2020). Specifiers and buyers of GUV products need accurate performance claims and data to deploy GUV technology safely and effectively. This CALiPER GUV Round 1 report demonstrates the significant education and training manufacturers and vendors still require to accurately test and report the performance of their GUV products. Further industry standards and guidelines may address testing limitations and improve test methods, product performance, and the accuracy of performance claims.

42 ENGINEERING↗

DOE OSTI2024

Performance Testing of a Moving-Bed Gasifier Using Coal, Biomass, and Waste Plastic Blends with Washed and Unwashed Legacy Coals and Other Waste Fuels to Generate White Hydrogen

The objective of this effort, primarily funded by the United States Department of Energy (DOE), and led by the Electric Power Research Institute, Inc. (EPRI), with support by Hamilton Maurer International (HMI) and Sotacarbo S.p.A. (Sotacarbo), has been to qualify coal, biomass, and plastic waste blends based on performance testing of selected fuel pellet compositions in a pilot-scale updraft moving-bed (UDMB) gasifier. The testing provided relevant data to advance the commercial-scale design of the moving-bed gasifier to be able to successfully use these feedstocks to produce hydrogen. In particular, the effects of waste plastics on feedstock development (i.e., blending and pelletizing) and the resulting products (i.e., syngas compositions, organic condensate production, and ash characteristics) are the focus. The gasifier used for testing is HMI’s moving-bed gasifier, which has been proven capable of gasifying nearly all coal ranks. It has also shown the ability in prior testing work to gasify wood chips (biomass). However, mixtures of these fuels with plastic wastes have not been prepared and gasified together. The three feedstocks were densified and pelletized by California Pellet Mill (CPM) to meet the feedstock size required by Sotacarbo’s 30mm ID UDMB gasifier, under contract to HMI. The technical tasks and results from this two-year research project included: (1) Feed Procurement and Preparation: Nine different tri-fuel pellets were prepared from varying compositions of fresh mined PRB coal, corn stover biomass, and car fluff waste plastics. Tri-fuel pellets were produced by CPM and shipped to Sotacarbo’s test facility in Carbonia, Sardinia, Italy. (2) Test Plan Development: A test plan was created to define the test runs to be performed. The test plan detailed the different UDMB gasification tests to be performed in Sotacarbo’s 12-inch ID pilot scale gasifier, the process monitoring instrumentation used, and the extractive samples recovered for analysis of the total gasification process mass and energy balance. (3) Gasifier Testing: Nine different gasification runs were performed in the pilot-scale gasifier at Sotacarbo using nine different fuel feedstock compositions generated from varying mixtures of PRB coal, biomass, and plastic wastes. The testing generated performance data on gasification reaction efficiency and performance, yielding relevant data for models used to scale up the gasifier design. This task also included work to refurbish and reassemble the pilot gasifier at Sotacarbo and perform a baseline 100% PRB coal run. (4) Data Analysis and Reporting: Review of the data, determination of figures of merit, and interpretation of the results are reported in the project’s final report, published in March 2024. The results show that all tri-fuel pellets gasified well and maintained structural integrity throughout the gasification process. The syngas generated can be shifted to hydrogen by using commercial syngas shifting technologies. (5) High Fidelity computational fluid dynamics (CFD) Simulation: The National Energy Technology Laboratory (NETL) team performed CFD simulations of the UDMB gasifier for two of the tri-fuel pellets gasified in Sotacarbo’s pilot scale gasifier. The kinetic mechanisms for the pyrolysis of each constituent, PRB coal, corn stover biomass, and waste plastics are based on thermogravimetric analysis performed by Sotacarbo. The gasification model was validated by comparing the predicted syngas composition at the exit of the gasifier with the measured syngas composition. In addition, the reactor’s measured internal temperature profile agreed well with the predicted internal reactor temperature profile. These results validate that the model can be used to predict the performance of the updraft moving bed gasifier for different feedstocks and operating conditions. This paper summarizes the results of the completed work in which the pelletizing procedure was validated to ensure the viability of the tri-fuel pellets for the gasification runs performed at Sotacarbo’s 30 mm UDMB gasifier. The gasification performance data from this series of nine runs will enable modeling of a full-scale HMI industrial scale gasifier supporting both combined heat and power, and Hydrogen production from coal (both fresh mined and legacy) combined with various biomass and waste plastics. Additionally, plans and progress on a follow-up project, being executed by the same project team, will be presented. In this project, a total of twenty (20) different feedstocks are being prepared from varying compositions of biomass (both woody biomass and corn stover) with a mixture of legacy coal waste, plastic waste, and refuse-derived fuel (RDF). The testing will provide information on gasification reaction efficiency/performance, yielding relevant data for models used to scale up the gasifier design to 50 megawatt electric (MWe) (equivalent hydrogen production). Tests will also be performed on a bench-scale fluidized-bed gasifier for comparison purposes. The results of this testing will be used to specify the range of feedstock blends that can be successfully gasified as well as quantify gasifier outputs based on specific blends.

08 HYDROGEN↗

DOE OSTI2024

FY24 Report on Water NSTF Testing: Lower Tank Inlet Piping Configuration

The following report serves as a summary of the accomplishments and testing results by the Natural convection Shutdown heat removal Test Facility (NSTF) experimental program over the past 12-month period. A major activity included reconfiguration of the chimney piping geometry which altered the discharge position into the tank from the original 50% tank height to a new lower position at a 10.9% tank height. This modification increases the volume of available coolant thus extending long-term operating capacity, however also results in a decreased liquid driving head which has the potential to reduce the natural circulation efficiency and decrease overall performance. Examination of the tradeoffs for these two configurations is an area of interest for RCCS designers and drove planned test activities this year. Ten matrix tests were performed in FY24, totaling 195 hours of heated operations and 12.9 MWh of electrical heating throughout the year, with eight classified as Accepted per NQA-1 and two classified as Trending. Testing prior to the facility reconfiguration examined the facility response to throttling at the tank inlet, demonstrating increased sensitivity of flow instabilities to throttling within the two-phase region when compared to throttling at the lower sensitivity single-phase inlet region. Testing after the chimney reconfiguration began with a baseline test at conditions of 80% inventory fill and prototypic thermal power input of 2.1 MW t . In addition to establishing a reference for nominal system behavior and performance, repeat testing with multiple subsequent runs demonstrated strong repeatability between tests performed at the same conditions in this new configuration. These tests also examined if influences would occur to system behavior with installation of new higher-resolution instrumentation within the upper chimney. In this critical region where boiling and flashing phenomena dominate, the test results provided confidence that the new instrumentation does not uncharacteristically influence the observed behaviors. An inventory parametric series was then initiated to examine system behavior with the new chimney configuration at six varying initial inventory levels, ranging from high 80% to low 20% fill. Two-phase oscillations with similar peak and mean flow rates were observed when comparing to the previous mid tank configuration. Generally, similar system response trends with inventory were also observed, such as two phase oscillations suppressing as inventories were lowered. However, in one absolute inventory comparison at the highest fill of 80%, flow oscillations saw a gradual growth over the 4 hours of two-phase operation in the new lower tank inlet configuration, opposite of the gradual dampening observed in the mid tank inlet configuration. This can be attributed to, in part, a greater hydrostatic head pressure above the two-phase discharge region where the boiling front is developed. Furthermore, the change resulted in greater sensitivity to liquid degassing phenomena during single-phase heating, causing loop instabilities to form which trigged moderate flow degradation during the period approaching saturation and boiling conditions. This behavior had been observed previously under some conditions but was a common occurrence in recent testing with the newer lower tank inlet configuration. Initial observations from these first data sets suggest an overall larger window of stability for the mid tank inlet configuration compared to the lower tank inlet. Lastly, the lowest inventory fill test was repeated over an extended testing window to examine depletion behavior. Natural circulation flow and effective heat removal performance were observed during most of the testing period; only after the tank became fully drained (0% fill) did flow stagnate and violent geysering events occur. This early observation confirms one relative advantage over the mid tank inlet configuration, which stagnated under comparable conditions at ~20% inventory remaining in the tank.

42 ENGINEERING↗

DOE OSTI2016

Special Form Testing of Sealed Source Encapsulation for High-Alpha-Activity Actinide Materials

In the United States all transportation of radioactive material is regulated by the U.S. Department of Transportation (DOT). Beginning in 2008 a new type of sealed-source encapsulation package was developed and tested by Oak Ridge National Laboratory (ORNL). These packages contain high-alpha-activity actinides and are regulated and transported in accordance with the requirements for DOT Class 7 hazardous material. The DOT provides specific regulations pertaining to special form encapsulation designs. The special form designation indicates that the encapsulated radioactive contents have a very low probability of dispersion even when subjected to significant structural events. The special form designs have been shown to simplify the delivery, transport, acceptance, and receipt processes. It is intended for these sealed-source encapsulations to be shipped to various facilities making it very advantageous for them to be certified as special form. To this end, DOT Certificates of Competent Authority (CoCAs) have been sought for the design suitable for containing high-alpha-activity actinide materials. This design consists of the high-alpha-activity material encapsulated within a triangular zirconia canister, referred to as a ZipCan, tile that is then enclosed by a spherical shell. The spherical shell design, with ZipCan tile inside, was tested for compliance with the special form regulations found in 49 CFR 173.469. The spherical enclosure was subjected to 9-m impact, 1 m percussion, and 10-minute thermal tests at the Packaging Evaluation Facility located at the National Transportation Research Center in Knoxville, TN USA and operated by ORNL. Before and after each test, the test units were subjected to a helium leak check and a bubble test. The ZipCan tiles and core were also subjected to the tests required for ISO 2919:2012(E), including a Class IV impact test and heat test and subsequently subjected to helium leakage rate tests [49 CFR 173.469(a)(4)(i)]. The impact-tile test unit contained a nonradioactive surrogate; however, the thermal test unit contained a radioactive source. This paper describes the regulatory special form tests and presents detailed impact and leak test results that demonstrate that the sealed source encapsulation designs satisfy the regulatory tests.

Martinez, Oscar A↗

DOE OSTI2020

An Accelerated Creep Testing Program For Nickel Based Superalloys

In order to improve the energy production within the United States, new materials are developed with the goal of increased functionality. These materials are required to perform at extreme pressures and temperatures (up to and beyond 200MPa and 650â??) as well as be sustainable beyond 105 hours. Before developing these materials into components for power generation, mechanical testing is required to better understand the behavior of these materials at various loads and temperatures. The longest of these tests is the conventional creep test which is a real time test that can beyond 105 hours depending on service conditions. Hundreds of specimens must be tested before being approved for use in power generation, but it is extremely difficult to rapidly qualify when each material is required to exceed 105 hours making testing through conventional means unfeasible. While others have designed accelerated testing methods to help qualify materials, many of these experiments use an empirical approach that requires test data to fit already existing creep data. In this study, an accelerated creep test called the stepped isostress method (SSM) is then applied to wrought Inconel 718 (IN 718) specimens. A test matrix is designed using deformation mechanism maps, time-temperature-transformation diagrams, and time-temperature-precipitation diagrams, in order to ensure that microstructurally, the accelerated creep tests fail similarly to the conventional creep tests. The data from the stepped isostress method is then applied to the McVetty and Sin-hyperbolic (Sinh) constitutive models in order to generate a creep master curve. The results from the model fits are then compared to the conventional creep data done from experiments as well as from the NIMS database for Inconel 718. Optical microscopy and scanning electron microscopy are used to compare the evolution of the microstructure from the conventional creep and accelerated creep tests. It is found that using the Sinh model on the accelerated creep data provides reasonable creep rupture predictions when compared to the scatter band of creep tests. It was also observed that the microstructural evolution between the SSM experiments and creep tests were very similar in terms of the volume fraction of precipitates and phases.

Mach, Robert↗