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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Sister Rod Destructive Test Results (FY19)

This report documents work performed under the Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-19ORO010201026 within work package SF-19OR01020102 and is an update to the work reported in M2SF-19OR010201028 and M3SF-19OR010201027. The document incorporates comments received with respect to the previous reports and incorporates additional test results. As a part of the US Department of Energy (DOE) Office of Nuclear Energy (NE) High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the “sister rods,” are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zircaloy-4 (LT Zirc-4), ZIRLO ® , and M5 ® . The DEs are being conducted to obtain a baseline of the HBU rod’s condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding will be tested to derive material properties. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and consolidation activities. This report documents the status of the ORNL Phase I DEs of 8 sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans. The DEs are performed using a phased approach, and the Phase 1 DEs being performed at ORNL include: • full-length rod heat treatments (FHT) of 3 selected sister rods to examine the effects of temperatures reached during dry storage preparation, • rod internal pressure and void volume measurements of the 3 FHT rods and 3 corresponding baseline rods, as well as 2 additional rods selected for depressurization/gas transmission tests and fatigue lifetime tests, • fission gas sampling and analysis, • depressurization and gas transmission tests, • rough segmentation of the selected rods for mechanical tests and rod characterization, • fuel sampling and burnup analysis, • metallography, • cladding total hydrogen measurements, • mechanical testing, including fatigue lifetime (CIRFT), four-point bend, axial tension, microhardness, ring compression, and burst tests. Table S-1 provides a summary of the DE status. The mechanical testing will be performed using fueled segments and is expected to complement previous and current mechanical test results using defueled cladding segments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Robustness of Asymptotic Optimal Accelerated Test Plans to Small-Sample Settings

Accelerated testing is a form of testing commonly employed in industry, especially at Sandia National Laboratories, in order to extract information about product lifetime behavior that cannot be acquired under normal operating conditions due to the extensive durations of the intended life- times. As of the writing of this report, there is no known guideline for the design of accelerated tests at the laboratories and yet there is an entire field in the statistical literature on the optimal design of accelerated tests covering a wide range of limitations and constraints available for reference. A potential roadblock to incorporating this knowledge is the common difficulty at Sandia of having severely limited data available for such testing. Nearly all of the methodology presented in the statistical literature is based on asymptotic or large-sample theory and so may not be appropriate for the testing conditions present at Sandia. This report investigates optimal accelerated test plans derived based on small-sample or exact methodology for both a basic test setting as well as a more realistic setting and compares the resulting test plans to those derived based on the large- sample methodology from the literature. It is shown that the large-sample test plans are actually quite robust to the presence of limited data and are generally more flexible than the small-sample test plans.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Technical and Programmatic Consequences of a Hypersonic Vehicle Flight Test Ban

The hypersonic vehicle (HV) is defined as any vehicle capable of traveling at or above five times the local speed of sound (Mach 5). Development of hypersonic flight has been pursued since the late 1950s, especially for reentry of missile warheads and returning orbiting spacecraft. Although considerable progress has been made since then, air- breathing and gliding hypersonic flight remains in the development and testing phase. Concerns have been raised, however, about continued HV flight tests because of HV's potential military use, which has the potential to encourage hair-trigger tactics that would increase crisis instability. In recent years, non-governmental organizations have called for a ban on HV flight testing for the primary purpose of halting continued development of the HV as a weapon. The emphasis on banning flight tests is due to its vital role in aerospace programs. Without it, significant technical uncertainties remain, creating safety and reliability issues that would discourage commercial development. Although the cost and time involved in flight tests have already spurred the use of other test methods, none can completely reproduce the flight test environment. Flight tests are important for reasons such as generating data unique to flight test environments to use in validating models and verifying ground test data, identifying unanticipated problems, and reducing risk. Commercial HV technology development today depends on advancements in four areas: propulsion; aerodynamics; critical guidance, navigation and control; and thermal management. This report summarizes the ramifications on HV technology development if flight tests were banned.

42 ENGINEERING↗
DOE OSTI2020

Iodine Waste Form Corrosion: PNNL Summary Testing Report

A test method to correlate the laboratory corrosion behavior of iodine-containing waste forms (IWFs) to a conceptual long-term corrosion models is required. Several candidate methods have been evaluated previously, and each method presents unique challenges and limitations. Dynamic leach tests have emerged as the preferred candidate test method to evaluate IWFs and parameterize an eventual long-term corrosion model. This report was prepared in support of this effort to define the test suite to evaluate the corrosion of IWFs, with the goal of defining and parameterizing a conceptual long-term corrosion model. A review was presented highlighting the high degree of variability between the IWF corrosion data that is available in the literature and the need for a standardized test suite. The targeted test suite is composed of a single interval screening test, a dynamic leach test, and post-test solid characterization of the monolithic IWF samples. The data presented serves two purposes: to expand the range of IWFs evaluated with the test suite to assess method variability and to further understand the corrosion of IWFs to best inform conceptual long-term corrosion model development. A series of IWF types were included in the testing, including hot isostatically pressed Ag-mordenite, silver functionalized silica-based aerogels, iodosodalites (hot isostatic pressing, sintered, glass bonded ceramic), iodoapatite, tellurite glass and glass composite materials.

42 ENGINEERING↗
DOE OSTI2021

UPF and Y-12 Criticality Accident Alarm System (CAAS) Radiation Testing Report

This report documents the experimental conditions and results for the testing of the Mirion Technologies CAAS-3S Criticality Accident Alarm System (CAAS) for the Y-12 National Security Complex (Y-12) and Uranium Processing Facility (UPF). The testing was conducted at the Godiva IV Burst Reactor at the National Criticality Experiments Research Center (NCERC) at the Nevada National Security Site (NNSS) during the week of January 11, 2021. The testing was conducted in accordance with 25774-QL-POA-JR00-00002-VDE-27.0, CAAS-3S Radiation Testing of Cabinet Components Test Plan. The tests subjected the CAAS-3S system to intense and short duration mixed neutron and gamma radiation fields and to high integrated radiation doses. The purpose of the test was to environmentally qualify various CAAS-3S components in a radiation environment. Successful completion of the tests was predicated on the collaborative efforts of several organizations. Sandia National Laboratory (SNL) and Lawrence Livermore National Laboratory (LLNL) provided dosimetry to estimate photon and neutron dose. Los Alamos National Laboratory (LANL) manages the NCERC facility, operated the Godiva IV reactor, and provided field support for CAAS-3S placement and communications interface for data acquisition systems supplied by Mirion Technologies. Throughout the test campaign, Y-12 and UPF personnel provided technical support in determining changes to the test schedule, as needed, interpretation of the results, and comparison between the measured dosimetry data and predicted results. Throughout the test campaign, Mirion personnel provided technical support in the operation of the CAAS-3S system itself.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Rapid, Approximate Multi-Axis Vibration Testing [Slides]

Sequential single-axis testing provides a poor approximation of service environments. Multi-axis vibration tests are not yet standard practice. How good do multi-axis tests need to be to replace single-axis testing in practice? Is the simplest possible multi-axis test close to single-axis testing in test quality? Can single-axis testing techniques provide a sufficient rapid approximation to a multi-axis test? How much can we improve multi-axis test quality by optimizing passive test fixture hardware instead of adding active excitation sources?

42 ENGINEERING↗
DOE OSTI2020

Analysis of Air-Purifying Respirator (APR) and Powered Air-Purifying Respirator (PAPR) Cartridge Performance Testing on Hanford Tanks BY-108 and BY-110 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 performance for APRs and PAPRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the headspaces of Hanford BY-108 and BY-110 waste storage tanks. 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 tests were conducted over two days from February 23-25, 2018, using headspace vapors from Hanford tanks BY-108 and BY-110. Headspace vapors from the BY-108 tank were fed to the PAPR respirator cartridge test stand, while vapors from the BY-110 headspace were fed to the APR respirator cartridge test stand. Both the APR and PAPR test stands were developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Two different cartridges were assessed on each tank. Multipurpose APR cartridges—SCOTT 7422-SD1 and SCOTT 7422-SC1 (SCOTT Safety, Monroe, North Carolina)—were assessed for vapors from BY-110 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)—also were assessed for vapors from BY-108 over the same two days 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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Assessment of Potential Dose and Environmental Impacts from Proposed Testing at the INL Radiological Response Training Range

This assessment uses screening-level models to calculate potential environmental impacts from proposed tests at the Idaho National Laboratory (INL) Radiological Response Training Range (RRTR) site. Proposed tests could be conducted using 11 different radioactive material types that include K2O, LaBr3, KBr, Cu, Zr, F, Ga, Ga2O3, NaNO2, Ga-68, and Tc-99m. The tests could potentially release radioactive material to the atmosphere and radionuclides and other contaminants to the soil, which could leach into the unsaturated zone and migrate to the aquifer. Atmospheric transport of radionuclides to potential human receptors and time-integrated air concentrations were calculated with a Gaussian plume model and three years of hourly meteorological data. Potential surface-soil impacts were calculated with the computer program mixing-cell model (MCM). Groundwater impacts were calculated with the computer programs MCM and GWSCREEN. Radiological doses from potential atmospheric releases were calculated for public receptors off the INL Site and for workers at nearby INL facilities. Results were compared to regulatory dose limits. Maximum potential groundwater concentrations were estimated in the aquifer below the NSTR site and compared to drinking water standards or risk-based screening levels for resident tap water. Soil concentrations were calculated and compared to risk-based screening levels for workers and potential future residents. All impacts were estimated assuming 12 tests are conducted annually using all 11 material types for a period of 15 years. This document provides the resources to enable a subject matter expert in the field of environmental assessments to replicate the modeling and calculations. The methodology and parameters are presented in the text. All electronic files, including computer-code input, output, executable files, batch files, scripts, and spreadsheet files are contained in a zip file that can be accessed by selecting “Additional Information” (select Native File) in the INL Electronic Document Management System (EDMS). It is highly unlikely the test scenarios evaluated in this ECAR will adversely impact human health based on comparisons of calculated dose and concentration against regulatory standards and risk-based screening levels. Conservative estimates of dose to workers and the public from atmospheric transport of possible radionuclide releases are far below federal radiation protection standards. Conservative estimates of potential contaminant concentrations in groundwater are less than federal drinking water standards or screening levels. Predicted radionuclide concentrations in surface soils are below risk-based screening levels, except for Ge-68 (material Ga-68) for the worker. The Ge-68 soil concentration can be made less than the worker PRG, if the number of annual tests using Ga-68 is reduced from 12 to 6. However, the sum of ratios still exceeds one because of the high K-40 ratio. If the EF of the worker (number of days the worker is in the contaminated testing area) is reduced from 225 days/yr (default value for full time worker) to 112 days/yr, the Ge-68 ratio is less than one and the sum of ratios is less than one. Actual radiation doses and groundwater and surface-soil concentrations are likely to be much less than those calculated because of the conservative assumptions and parameters employed in the modeling. For example, atmospheric-transport calculations assume the entire inventory of each material type is readily released to the atmosphere and no plume deposition, depletion, or radioactive decay occurs during transport. The calculations also assume the same meteorological conditions (e.g., wind velocity, wind direction, stability class) that produce the maximum 95th percentile concentration (i.e., concentration representing the 95th percentile of a distribution of concentrations derived from 3 years of hourly meteorological data) at each receptor location are the same for all 12 tests during the year, and each receptor is assumed to be present during all 12 tests. The surface-soil assessment assumes the entire inventory of each test is deposited in the top 5 cm of soil. No atmospheric dispersal is assumed, and the radionuclides are subject only to leaching and radioactive decay. The groundwater-pathway modeling is conservative in that it is one-dimensional in the unsaturated zone (no lateral spreading/dilution) and assumes the entire inventory of contaminants infiltrates into the ground at the same location for every test. This is especially conservative for particulate radionuclides because they would have to dissolve or corrode first and some would be dispersed into the atmosphere. The groundwater receptor is also assumed to consume water directly from a hypothetical well positioned in the location of maximum concentration. In addition, conservative degradation rates were used, and volatilization was not considered for the nonradioactive chemical

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

ChargeX Prescribed Testing Program at CharIN June 2024 Testival: Outcomes & Future Recommendations

In June 2024, the ChargeX Consortium developed an optional prescribed testing program for electric vehicle (EV) and electric vehicle supply equipment (EVSE) manufacturers that attended the CharIN Testival as testers. There were two driving purposes of this program; to introduce a new hybrid approach to testing events with both ad-hoc and prescribed testing offered, and to demonstrate the test cases and structure effectiveness of the EV-EVSE Interoperability Test Plan (EEITP) document developed within the ChargeX Testing Task Force. This program contained eight test scenarios to be performed during the final 30-minutes of a 90-minute testing slot with details like purpose, setup, pass criteria, etc. included within a written test plan document. A $2,000 rebate was offered to those who participated, and a ChargeX moderation force was present to collect EV and EVSE meta data, testing meta data, and testing results.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2022

Mechanics of the Ring Tension Test (RTT): A Finite Element-based Investigation

The ring tension test (RTT) is an experimental method for determining mechanical behavior in a material’s circumferential or hoop direction. It is a crucial test for testing anisotropic materials with tube geometry, such as nuclear fuel cladding or irradiated pipes. Several RTT configurations exist, each with their own advantages and disadvantages. However, this test is significantly more complex than traditional tensile testing and can be especially sensitive to small differences and inconsistencies in the test setup and geometry, ultimately affecting the derived mechanical properties. Previous research has focused on method development, and little work has been done on understanding the subtle differences between an ideal test and experiments, specifically when the tests are performed on highly irradiated materials in hot cells. In this work, a finite element-based investigation of the RTT is conducted. Two promising test configurations are investigated, comparing their ability to determine accurate material strengths through plastic deformation. Several non-ideal conditions and uncontrollable effects which are likely to occur during experimental testing such as machining tolerances, variations of specimen geometry from nominal dimensions, rotation of specimens and fixturing, and other test setup discrepancies are studied. The sensitivity of measured strengths to these conditions is presented. A mechanics-based approach to describing and correcting raw data to determine actual strengths is also included for one of the configurations, resulting in a robust correction method with highly accurate material strength measurements. Based on these analyses, the hemicylindrical mandrel configuration is recommended with a gauge region oriented at a 45° angle.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

The Hanford Waste End Effector Phase II Test Report

This report describes the Phase II testing program of the Hanford Waste End Effector (HWEE) at the Washington River Protection Solutions Cold Test Facility (CTF). Phase II testing builds on the completed Phase I development testing (PNNL-26856) to address some key actual waste challenges as outlined in the Phase II test plan (PNNL-27422). This testing demonstrated retrieval of a mixed layered simulant and retrieval of uneven simulant profiles like that found near tank walls. A separate demonstration using only the HWEE conveyance jet pump assessed the lifting capabilities of this device for transporting waste out of an underground tank. The Pacific Northwest National Laboratory (PNNL)-designed robotic gantry system and associated systems (e.g., tanks, high-pressure water delivery and valves, instruments, etc.) used at CTF during Phase I were modified and upgraded to implement the Phase II test program and measure various process parameters. The tests demonstrated retrieval of water and Kaolin (sludge) simulant to determine pumping rate, dilution factors, and screen fouling rate. This report will first describe the test equipment, instrumentation, simulant preparation procedures, and testing methodology. Testing results presented include HWEE water usage versus operating pressure of modified components from Phase I, simulant characterization data, dilution ratios, and retrieval rates. Reaction forces and other factors that would affect the remote operation of the HWEE will also be presented. Results of the separate jet pump vertical lift demonstration to a fixed height of at least 35 feet are also presented.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.

Crack Growth↗
DOE OSTI2018

AGR-5/6/7 Irradiation Experiment Test Plan

This document presents the test plan for the AGR-5/6/7 irradiation experiment, the combined fifth, sixth, and seventh irradiations of the United States Department of Energy Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The combined irradiation tests will consist of fuel qualification tests (AGR-5/6) and a margin test (AGR-7) for the industrially produced uranium oxycarbide (UCO) tristructural isotropic (TRISO) coated particle fuel developed by the AGR Program. The objective of the AGR-5/6/7 irradiation experiment is to establish the performance of UCO TRISO fuel fabricated using engineering-scale equipment in a production environment under and beyond normal operating conditions. Specifically, the goals of the AGR 5/6/7 irradiation experiment are: 1. Irradiate reference design fuel containing low-enriched UCO TRISO fuel particles to support fuel qualification. 2. Establish the operating margins for the fuel beyond normal operating conditions. 3. Provide irradiated fuel performance data and irradiated fuel samples for post-irradiation examination and safety testing. 4. Support the refinement of fuel performance and fission product transport models with online, post-irradiation examination and safety test data. In order to achieve the test objectives, the AGR-5/6/7 experiment will be irradiated in the northeast flux trap position of the Advanced Test Reactor at Idaho National Laboratory. The larger diameter of the northeast flux trap provides greater flexibility for test train design, significantly enhancing the capability for the combined irradiations. The test train contains five separate and independently controlled and monitored capsules. The middle capsule contains the AGR 7 margin test fuel. The two bottom and two top capsules contain the AGR 5/6 fuel. Each capsule contains a variable number of fuel compacts filled with UCO TRISO particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Destructive Examination of a FeCrAl-UO 2 Irradiation Test

Several destructive postirradiation examinations were performed on an irradiation specimen that coupled an early iron-chromium-aluminum (FeCrAl) candidate alloy cladding with UO2 pellets. This irradiation test was designed to investigate the early life performance and compatibility of the FeCrAl-UO 2 cladding-fuel system under prototypic light water reactor neutronic conditions. Additionally, these tests were expected to provide neutron irradiated samples for severe accident testing of this fuel system. The irradiation studied in this work was part of the ATF-1 series of drop-in style irradiations performed in the Idaho National Laboratory Advanced Test Reactor. The rodlet studied in this work is one of three similar rodlets irradiated in ATF-1 that had approximately 7.6 cm of cladding machined from a wrought FeCrAl alloy and fueled with a 6.1 cm stack of UO 2 pellets. The outer diameter of the cladding was ~0.94 cm and the inner diameter was ~0.83 cm. After irradiation and non-destructive examination in Idaho this rodlet was shipped to the Oak Ridge National Laboratory hot-cells for further examination. The irradiated rodlet was sectioned into several samples for microstructural, micromechanical, and severe accident testing. During sectioning, it was noted that the fuel was not firmly bonded to the cladding and could be readily removed from small cladding slices. Microstructural characterization of fuel cross sections also revealed no significant interaction between the fuel and the cladding. Samples were also prepared for microhardness testing. To prepare for high temperature oxidation testing, the fuel was dissolved from segments of the cladding. High temperature oxidation testing of cladding segments was performed at 1200°C and 1300°C in a steam environment. Comparisons between the oxidation of this FeCrAl alloy in its neutron irradiated state, as-fabricated state, and the oxidation of Zircaloy-2 are made. The oxidation testing will be followed by ring compression testing to evaluate ductility. The microstructure of the samples after oxidation and ring compression testing will also be analyzed.

Harp, Jason↗
DOE OSTI2023

Recent Radioactive Material Package Testing Experiences at Oak Ridge National Laboratory

Oak Ridge National Laboratory (ORNL) was a pioneer in the testing of radioactive material shipping packages and has been performing such tests for over 50 years. Currently, except for Type B thermal tests, testing is performed at the purpose-built Packaging Research Facility that is located at the National Transportation Research Center - a U.S. Department of Energy User Facility. Packages used for the transportation of Type B quantities of radioactive materials must be capable of meeting normal conditions of transportation (NCT) requirements and hypothetical accident conditions (HAC) requirements as defined in Title 10 Code of Federal Regulations (CFR) Part 71. Evaluation of package designs to these criteria typically involves a combination of analysis and physical testing. Required physical tests include free drop, crush, puncture, penetration, compression, vibration, water spray, water immersion, and thermal. The Oak Ridge National Laboratory (ORNL) Packaging Evaluation Facility (PEF) at the National Transportation Research Center (NTRC) provides an inclusive testing capability for a wide variety of packages needed for shipment of radioactive materials. Since PATRAM 2019, ORNL has performed several sets of tests including testing of two designs of the Defense Program Package (DPP-3 and DPP-1) and several Special Form Capsules. This paper provides an overview of the testing performed as well as a summary of advancements made to the ORNL package testing program during that time.

Martinez, Oscar↗
DOE OSTI2016

Analyses of SPERT-CDC test 859 by FEMAXI-7 and RANNS codes

In the current Japanese regulation concerning fuel safety of water-cooled reactors, the criterion of fuel failure due to pellet-cladding mechanical interaction (PCMI) in a burnup range of 25-40 GWd/t is determined substantially based on the result of SPERT-CDC test 859 (SPERT859). However, there is room for discussion if the result of SPERT859 test rod with a burnup of 31.8 GWd/t represents the failure behavior of the fuel rod for commercial light water reactor because it was reported in some literatures that the cladding of SPERT859 test rod corroded excessively during pre-irradiation. On the other hand, the oxide thickness on the cladding of SPERT859 test rod was not reported in the data of its post-irradiation examinations (PIEs) after pre-irradiation, and the probability of the excessive corrosion on the cladding of the test rod and the effect of such excessive corrosion on the fuel enthalpy at failure of the test rod has not yet been clarified. In this study, the oxide thickness of the cladding formed on the cladding outer surface of SPERT859 test rod and its fuel enthalpy at failure due to PCMI under this corrosion condition were analyzed by using fuel performance codes FEMAXI-7 and RANNS. From the analysis results of FEMAXI-7, it was found that the excessive corrosion of the cladding of the test rod should be assumed to reproduce the fuel outer diameter measured in PIEs after pre-irradiation. The fuel enthalpy at failure of the test rod was calculated by using RANNS, and the calculated fuel enthalpy at failure was relatively close to the value of SPERT859 test rod. These results suggested that the fuel enthalpy at failure of SPERT859 was affected by the excessive corrosion on the cladding of the test rod and was likely lower than that of the typical fuel for light water reactors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Evaluating the Feasibility of Using Strain Measured During Sinusoidal Rate Pumping Tests to Characterize an Aquifer

Pumping tests with sinusoidal variation in pumping rate have been proposed as a method for improving aquifer characterization. These tests can interrogate a larger aquifer volume than slug tests and they can be more sensitive to small variations in drawdown. Current methods of using sinusoidal variations of rate are based on measuring pressure signals from the reservoir or aquifer, which requires access to monitoring wells. An alternative approach has been developed that measures the strain in the vadose zone instead of pressure in the reservoir. An instrument has been developed at Clemson University that can measure small strains using optical fiber sensors, and we propose that it can be used to measure strain during sinusoidal pumping tests. The objective of this project is to evaluate the feasibility of measuring and interpreting the vertical strain in the vadose zone generated during a sinusoidal rate pumping tests to estimate aquifer and/or overburden properties. Pumping tests were performed at a site with a water table at 9 m and saturated saprolite from 9 to 30m depth. Sinusoidal pumping tests in the saprolite used different pumping amplitudes and periods while pressure was measured at three piezometers and vertical strain was measured in the vadose zone from 1 to 7 m depth. The delay time of the pressure increased linearly with distance, and it increased with the square root of the period, while the amplitude of the pressure decreased with distance and increased with amplitude. Vertical strain in the vadose zone varied with the same period as the pumping rate and the delay times were similar to those for pressures measured at similar radial distances. Delay times for the strain increased with an increasing propagation of the strain field. The hydraulic diffusivity was estimated using delay times for pressure and strain, and the results were compared with the values estimated from a constant-rate pumping test. The hydraulic diffusivity from the constant rate test was 0.039 m2/sec, whereas it was 0.053 (±0.05) m2/sec and 0.035 (±0.03) m2/sec from the pressure data and 0.08 (±0.022) m2/sec and 0.035 (±0.001) m2/sec from the strain data from sinusoidal tests of various periods and amplitudes. There is no statistical difference among these values. This study indicates that strain measurements in the vadose zone can be used to estimate aquifer properties during sinusoidal pumping tests.

Blais, Riley↗
DOE OSTI2026

A Graphical User Interface for the Deep Underground Neutrino Experiment Robotic Test Stand

In preparation for DUNE, Fermilab along with six other institutions are testing cold electronics for quality control before components placed in the far detector. We test them by using a robotic arm that places these chips into sockets on a computer board that will test their functionality. Up until now, the chips have been tested using a command line script that drives a state machine to conduct tests step-by-step. In order to lower the skill barrier to conduct tests and to speed up the quality control process, I was tasked to create a graphical user interface that would allow users to use buttons, text boxes, and drop-down menus to input information and tell the testing state machine how to operate. I had to learn about the Python package Tkinter to start the process of widget placement. I further developed a pause feature unused in the previous command line script that would allow the user to shut down testing gracefully, bring the robotic arm to go back to ground state, and go forward or backward a step in the testing process. After completing the basic functionality of the GUI, I started testing production chips with the GUI to debug. Some issues were found, which required me to further develop parts of the inherited state machine code. The code for the GUI has now been pushed into the copy the DUNE/FD_CE git repository and will soon be merged with the official DUNE/FD_CE repository so that the other institutions testing DUNE cold electronics can use and expand upon it.

Gutierrez Villanueva, Jaziel [Fermilab]↗