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At least 865 records · Page 48

Hydrogen Storage Tests Phase 2: Advanced Cryocompressed Hydrogen Delivery and Onboard Storage (CRADA No. TC02207 Final Report)

Thin-lined, high fiber fraction vessels have been designed, built and tested for producing cryogenic pressurized hydrogen storage with weight (9+% hydrogen weight fraction) and volume ( 50+ grams of hydrogen per liter of system volume) onboard storage performance. Long-term vessel durability was tested and partially demonstrated to 456 cycles. Other accomplishments included cycle testing a vessel provided by BMW 1000 times to demonstrate durability; and building and testing a facility for efficient testing of vessels and hydrogen components.

42 ENGINEERING↗

BOPTEST: Initial FRP Test Case Development

To support an emulator model development of Building Optimization Testing (BOPTEST) framework, a team at the US Department of Energy’s Oak Ridge National Laboratory developed a medium-sized office building written in Modelica to serve as a reference test case. The reference test case was modeled based on an actual research platform that represents a typical medium-sized office building. This test case would encourage model-based design and testing of new controllers in which a developer first prototypes and debugs on a model before implementing in the real world. The developed preliminary model and control parameters will be further calibrated and tuned with actual measured data from experiments

42 ENGINEERING↗

SECARB Semi-Annual Field Review Reports - Early Test

The Southern States Energy Board (SSEB) managed the overall Southeast Regional Carbon Sequestration Partnership (SECARB) Phase III project and The Bureau of Economic Geology (BEG), at the University of Texas at Austin, managed all activities associated with the Early Test field site near Cranfield, Mississippi (the Early Test or Cranfield Project). BEG was assisted in the management of site activities by Sandia Technologies (Sandia). BEG and Sandia implemented reporting procedures that included monthly and quarterly technical progress reports, as well as periodic reporting on key field activities. Subsequent sections of this report catalog the field activities and are organized in a semiannual chronology. The Cranfield Project was located approximately 12 miles east of Natchez, Mississippi. Denbury Onshore, LLC, began operating a commercial CO 2 flood of the field (using the subsurface injection of CO 2 for EOR) in June 2008. Carbon dioxide from the Jackson Dome, a natural source near Jackson, Mississippi, was delivered to the Cranfield oilfield via pipeline. Beginning in October 2008, the SECARB Early Test team characterized the surface and subsurface of the Cranfield site. During the field test CO 2 was injected into the lower Tuscaloosa Formation, a regionally extensive saline formation with the potential to hold millions of tons of CO 2 emissions. By January 2015, the team injected more than 11 million metric tons of CO 2 and monitored a total stored CO 2 mass of 5,326,975 metric tons. SECARB’s study operations occurred in four integrated research program areas within Cranfield field: (1) the High Volume Injection Test area (HiVIT); (2) the Detailed Area of Study (DAS); (3) the Geomechanical area; and (4) the near-surface observatory, also called the “P-site.” Carbon dioxide injection activities were conducted at the HiVIT and the DAS. Figure 1.1-1 provides a depiction of the DAS.

01 COAL, LIGNITE, AND PEAT↗

Testing to Evaluate Processes Expected to Occur during MSR Salt Spill Accidents

Obtaining a license for a new nuclear reactor requires the identification and assessment of the potential consequences of specified accident scenarios, which are achieved using accident progression models. Those models need to be parameterized and validated using experimental data, but existing experimental data addressing processes relevant to molten salt reactor accidents are sparse. Specifically, experimental data that quantify the sensitives of important processes to the initial conditions of the spill, the ambient environment, and the containment features are needed to parameterize individual process models. Integrated experiments that simulate accident scenarios are also needed to provide data for model validation, but these experiments will require the use of proven methods to quantify the processes under evaluation. The overarching objectives of this work are to develop the methods for simulating the targeted processes, to determine the effectiveness of the methods in producing the data required for model development, to generate data that can be used to parameterize individual process models, and to provide key insights into the behavior of spilled molten salt that should be considered in models. Experimental methods were designed to quantify aspects of individual processes expected to occur during or after a molten salt spill accident that will affect the fate of spilled molten salt and the radionuclides within. These processes include 1) molten salt spreading and heat transfer, 2) molten salt flowing and freezing in tubing, 3) stainless steel corrosion kinetics in molten salt, and 4) molten salt splashing and aerosol generation. The initial tests described in this report were conducted using eutectic FLiNaK to demonstrate the test methods, the data that are generated, and the analyses of the data to derive values needed for modeling. The primary variables that were tested include initial salt temperature and the presence of volatile surrogate fission products (e.g., cesium and iodine). The developed methods are shown to be effective in quantifying the desired processes and can be applied to study more complex salt compositions of interest to molten salt reactor developers, a wide range of environmental conditions of interest to modelers, and additional variables relevant to salt spill accidents. The developed methods and insights gained from laboratory tests can also be incorporated in future large-scale integrated tests used to simulate molten salt spill accidents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Report on Field Test at INL Cask Farm of Single Detector Fast Neutron Spent Fuel Cask Verification System

Detecting diversion of spent fuel elements in dry storage casks is challenging due to the thick shielding used in cask construction. Measurements on top of the cask to map the underlying arrangement of the fuel elements and looking for anomalous changes over time has proven difficult to achieve using gamma rays due to the high scattering and attenuation from the thick steel structure, weakening information on the present or absence of fuel bundles. Simulations and laboratory experiments suggest that the high-energy neutron flux (>200 keV) measured directly above each fuel bundle is sufficient to produce a position map that enables detection of the present or absence of fuel bundles, and therefore diversion of a spent fuel bundle. A single-detector spent-fuel monitoring technique based on this principle was development at the Lawrence Livermore National Laboratory (LLNL). The INL Cask Farm in the INTEC technical area at Idaho National Laboratory (INL) offers the capability to test this technique on an MC-10 storage cask which has a distribution of full and empty fuel positions. An experimental test plan for the single-detector verification system was developed in consultation with INL personnel to be completed in FY2021. Due to travel advisories related to COVID-19, the experimental test plan was adapted to enable INL personnel to carry out the measurements in consultation with LLNL personnel following shipment of the LLNL system to INL. Field test measurements of the single detector verification system were successfully carried out at the INL cask farm on September 7-9, 2021. This document summarizes results of the field test.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of Real-Time, In-pile Creep Test Rigs for Characterizing the Structural Materials of Nuclear Components

New and improved materials are being considered for supporting both existing and next-generation nuclear reactors. Reactor materials can significantly degrade with time, thus limiting or altering their properties in harsh reactor environments. To accurately understand such material degradation, real-time data obtained under prototypic irradiation conditions are required. In particular, understanding the creep behavior of materials exposed to irradiation and elevated temperatures is essential for safety concern evaluations. To provide these capabilities, Idaho National Laboratory (INL)’s High Temperature Test Laboratory (HTTL) developed several instrumented test rigs for obtaining real-time data from specimens in well-controlled pressurized-water reactor (PWR) coolant conditions at the Materials Test Reactor. This technical report focuses on INL’s efforts to evaluate and enhance the former creep test rig prototype that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig can detect changes in the length of creep specimens, which is useful for measuring thermal expansion and creep deformation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

IHE DDT Qualification Tests (FY2020-FY2021)

A DDT tube test was designed to meet all requirements established in the new DOE IHE qualification standard. Ten tests were successfully executed with PBX 9502 as a demonstration of the design, assembly, heating system, and methods, and to establish a baseline set of test results with an existing IHE. The explosive response in all tests was low violence (as anticipated). This report contains detailed design drawings, assembly instructions and procedures for future execution of the qualification test series, and makes recommendations for minor improvements.

42 ENGINEERING↗

Attenuator, PC-2.92mm, Fixed (Proficiency Test Report; PT ID Number 6679665)

The PSL has reviewed the documentation and data provided by NNSS–Livermore Operations with respect to this proficiency test. This proficiency test was performed to assess NNSS–Livermore Operations’ ability to perform scattering parameter calibrations. The level of documentation was satisfactory. On 12/28/2021, NNSS–Livermore Operations reported the data for the proficiency test conducted on the attenuator. NNSS–Livermore Operations performed this proficiency test using an Anritsu vector network analyzer, an electronic calibration module, and verification kit. The PSL used a Keysight vector network analyzer and mechanical calibration kit. The PSL results included in this proficiency test report were taken on June 23, 2020.

42 ENGINEERING↗

Attenuator, PC-2.92mm, Fixed (Proficiency Test Report: Document # 6679664_11752120)

The PSL has reviewed the documentation and data provided by NNSS–Livermore Operations with respect to this proficiency test. This proficiency test was performed to assess NNSS–Livermore Operations’ ability to perform scattering parameter calibrations. The level of documentation was satisfactory. On 5/19/2020, NNSS–Livermore Operations reported the data for the proficiency test conducted on the attenuator. NNSS–Livermore Operations performed this proficiency test using an Anritsu vector network analyzer, an electronic calibration module, and verification kit. The PSL used a Keysight vector network analyzer and mechanical calibration kit. The PSL results included in this proficiency test report were taken on June 23, 2020.

47 OTHER INSTRUMENTATION↗

Analysis of Transient Groundwater Flow Near the Cheshire Test at Pahute Mesa, Nevada National Security Site

This report focuses on analysis and determination of the causes of transient water levels measured between 1964 and 2017 near the Cheshire underground nuclear test, which was detonated on February 14, 1976 in Area 20 (USDOE, 2015) of the Pahute Mesa (PM) testing areas of the Nevada National Security Site (NNSS) located in southern Nevada (Figure 1). Transient water levels near Cheshire indicate influences of a water-production well and yet-unexplained water level rise attributed to testing effects (Jackson and Fenelon, 2018; Jackson et al. 2021). Transient groundwater flow is relevant to the implementation of the hydrologic source term (HST), which prescribes the quantity and release rates of radionuclide source terms to be applied at individual test locations in assessment of regional-scale radionuclide transport (Carle and Tompson, 2021; Carle et al., 2021), as previously investigated for NNSS testing areas at PM (Pawloski et al., 2001), Frenchman Flat (Carle et al, 2007), Yucca Flat (Mcnab, 2009; Tompson, 2009), and Rainier Mesa (Carle, 2011; Tompson et al., 2011).

54 ENVIRONMENTAL SCIENCES↗

SAS4A/SASSYS-1 Validation with the EBR-II Tests SHRT-17 and SHRT-45R

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

2022 Los Alamos National Laboratory Test and Analysis Abstracts

The abstracts included in this document are: Responsiveness, Modernization, and Radiography Test at the Centrifuge Test Facility; LANL Hostile Blast Testing Capability Development; Digital Image Correlation Measurement of B61 Center Case Fragmentation Test; and Collaborative Analysis Testing Techniques (CATT) Phase 1.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

Analysis of Transient Groundwater Flow Near the Cheshire Test at Pahute Mesa, Nevada National Security Site

This report focuses on analysis and determination of the causes of transient water levels measured between 1964 and 2017 near the Cheshire underground nuclear test, which was detonated on February 14, 1976 in Area 20 (USDOE, 2015b) of the Pahute Mesa (PM) testing areas of the Nevada National Security Site (NNSS) located in southern Nevada (Figure 1). Transient water levels near Cheshire indicate influences of a water-production well and yet-unexplained water level rise attributed to testing effects (Jackson and Fenelon, 2018; Jackson et al. 2021). Transient groundwater flow is relevant to the implementation of the hydrologic source term (HST), which prescribes the quantity and release rates of radionuclide source terms to be applied at individual test locations in assessment of regional-scale radionuclide transport (Carle and Tompson, 2021; Carle et al., 2021), as previously investigated for NNSS testing areas at PM (Pawloski et al., 2001), Frenchman Flat (Carle et al, 2007), Yucca Flat (Carle et al., 2009; Mcnab, 2009; Tompson et al., 2009), and Rainier Mesa (Carle, 2011; Tompson et al., 2011).

54 ENVIRONMENTAL SCIENCES↗

Witness the Trinity test through Lab artifacts

July 16 marks the 77th anniversary of the Trinity test, conducted in a desert in New Mexico. The test subject, an atomic bomb called The Gadget, was successfully detonated from a 100-foot steel tower. This event marks the commencement of the Atomic Age, a new era where fission capabilities could be employed for national security purposes. Shortly after the Trinity test, two Los Alamos-created atomic weapons were released above Japan, helping to end the world’s bloodiest conflict just weeks later. “Trinity was one of the greatest scientific experiments ever,” said NSRC Senior Historian Alan Carr said. “Los Alamos scientists changed the world forever on that day. Not only was it the dawn of the Atomic Age, but also the beginning of the Lab’s eight decades of cutting-edge science and its national security charge.” To preserve this event, and to continue to learn more about this critical moment in history, the National Security Research Center (NSRC) curates a collection of photographs, films, notes, unclassified artifacts and numerous other materials related to the science of the test. Notably, the collection includes a novel material that formed at the site, trinitite, and artifacts from one of the intriguing scientists present at the test, Enrico Fermi.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

De-Energized Line Testing Interface (DELTI) Report

To guarantee the safety of power system operations, it is crucial to understand a de-energized power line’s status before restoring power. It is hypothesized that the Grid Resonance Probe (GRP) and De-Energized Line Testing Interface (DELTI) can be utilized to identify a faulted distribution line. Therefore, the project team is targeting development of a methodology to identify faulted distribution lines by using the DELTI system. In this R&D task, two field tests on Plum Island have been conducted, including DELTI tests with multiple configurations and fault scenarios. This report provides the analysis results of distribution line fault identification from the two field tests. In the remainder of this report, Section 1 is the introduction of the DELTI system, Section 2 provides the analyses of the two field tests in June and October of 2021, respectively, Section 3 discusses the existing line fault detection devices on market and how they compare to DELTI and Section 4 includes the lessons learned and recommendations for future work.

24 POWER TRANSMISSION AND DISTRIBUTION↗

Strain gauge for testing microreactor hexagonal core blocks in the Single Primary Heat Extraction and Removal Emulator

To support the development and deployment of microreactor technologies, experiments that help verify and validate reactor systems and components are performed at non-nuclear test facilities. The Single Primary Heat Extraction Removal Emulator at Idaho National Laboratory is one of these test facilities and was used to monitor a test article that has a prototypic geometry of a heat-pipe cooled microreactor core block. To collect crucial temperature and strain data during testing, temperature and strain sensing fiber optics were embedded to the surface of the test article using an ultrasonic additive manufacturing technique. To support and provide benchmark strain data for the embedded sensors, a commercial resistive strain gauge was attached. This report will discuss the results from the deployment of the commercial strain gauge which includes the setup/attachment strategies, data acquisition, and analysis of the strain data.

36 MATERIALS SCIENCE↗

Research and Test Reactor Fuels

PRO-RR is the research reactor focused program element of the broader Proliferation Resistance Optimization program (PRO-X) under the National Nuclear Safety Administration (NNSA) in the U.S. Department of Energy (DOE). PRO-X provides a framework for integrating proliferation resistance in nuclear system designs to minimize weapons usable nuclear materials (WUNM) production and diversion pathways while optimizing systems performance for peaceful use missions. PRO-RR applies the PRO-X mission objectives to research reactor system design. This document serves as one of the foundational documents for the PRO-RR-Fuel System Design technical team by documenting current research reactor fuels usage. The PRO-RR-Fuel System Design technical team consists of subject matter experts from Argonne National Laboratory (Argonne) and Savannah River National Laboratory (SRNL). In order to determine the preferred fuel of use in upcoming research and test reactors to optimize proliferation resistance, performance, and safety, it is useful to assess the fuels that have been used in the past, or are currently in use. This report reviews the historical and current fuels used in research and test reactors to inform future fuel selection. Chapter 2 discusses the low-enriched uranium (LEU) fuels currently in use in terms of thermal power level and utilization of the reactor. Chapter 3 summarizes the fabrication processes for common fuel types. Chapter 4 discusses in detail the fuel types in use in research and test reactors. A review of the cladding types in use is presented in Chapter 5, and a historical review of research and test reactor fuel fabricators is presented in Chapter 6. The data collection strategy used the International Atomic Energy Agency (IAEA) research reactor database [1] as a starting point. Information on the fuel used was gathered on research reactors (other than critical assemblies) that were listed as operational, planned, or in temporary shutdown in the IAEA database. Data on the fuel type, geometry, enrichment, uranium loading, cladding type, and fabricator were collected for each of the reactors available in the public domain. Sources of data included conference papers, journal articles, and facility and fabricator websites. Data on research reactors operating on LEU fuels are presented in Appendix A, while Appendix B presents data collected on all reactors at the time of publication of this report. Appendix C presents data collected on reactors that were part of the M3 research and test reactor conversion program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

SRPPF Aqueous Recovery System Sorbent Testing

This study details preliminary testing of 7 inorganic sorbents for the solidification of the surrogate SRPPF aqueous recovery system liquid excess material of 4M sodium nitrate. Sorbents were subject to various mixing methods to incorporate sodium nitrate solution including static, rotating, and in situ paddle mixing. Paddle mixing was found to be the most effective mixing technique, while simple static tests were also successful, though only for some of the sorbents. Setting tests were performed to ensure that the sorbent sequestered the liquid without release through a conical paint filter. Recommended set times ranged from near immediate to 7 days. The set mixtures were investigated further using ultra-centrifugal liquid release testing, thermogravimetric analysis, and visual microscopy to understand the mechanism by which the sorbent sequestered the liquid. Most sorbents seem to incorporate liquid by sorbing it onto and into the sorbent particles, while Drierite ® , anhydrous calcium sulfate, created a crystal structure that bound the surrogate liquid. Overall, four sorbents are recommended for continued testing: Aquaset II, Aquaset IIG, Drierite ® , and Spill-X-A.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗