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LCLS-II-HE Cryomodule Testing at Fermilab

22 Linac Coherent Light Source II (LCLS-II) cryomodules were successfully tested at the Cryomodule Test Facility (CMTF) at Fermilab. Following the completion of the LCLS-II testing program, CMTF has shifted to testing cryomodules for the LCLS-II High Energy upgrade (LCLS-II-HE). The first LCLS-II-HE cryomodule, the verification cryomodule (vCM), was successfully tested and verified the readiness of LCLS-II-HE cryomodule testing at CMTF, and production cryomodule testing has begun. Presented here are the production cryomodule test acceptance criteria, testing plan, and cryomodule test results so far.

43 PARTICLE ACCELERATORS↗

Material Testing of Coated Alloys in a Syngas Combustion Environment Year 6 - Activity 1.13 - Development of a National Center for Hydrogen Technology

Modifications were made to the inlet of the existing Energy & Environmental Research Center (EERC) thermal oxidizer to accommodate side-by-side coupon holders for exposure testing. Two 5-day tests with over 200 hours of total exposure time were completed. The first week of testing was conducted in enriched air-blown mode, with coupon temperatures ranging from 128° to 272°F. Carbonyl sampling was conducted, but it was discovered after the fact that the methodology used was producing very low recoveries of iron and nickel carbonyl. Therefore, the data generated during this week of testing were not considered accurate. The second week of testing was conducted in oxygen-blown mode, with coupon temperatures ranging from 220° to 265°F. Two improved methods were used to measure carbonyl concentration during this week of testing. These methods produced results closer to equilibrium calculations. Since both weeks of testing mostly produced a product gas with approximately 15%–18% carbon monoxide, it was felt that actual carbonyl concentrations for Week 1 should be very similar to those measured during Week 2. The revised carbonyl sampling methodology used during the second week of testing greatly improved the recovery of iron and nickel carbonyl in the sample. Even though the sampling results obtained from the first week were inaccurate, the results from the second week can be used as an estimate for the periods during which the gasifier was operating under similar conditions and producing similar product gas compositions. Specifically, Test Periods 2 and 3 from the first week were similar to the conditions run during the second week. For a product gas containing roughly 15%–18% CO and a coupon temperature of approximately 220°–270°F, the nickel carbonyl concentration should be about 0.05–0.1 ppm and the iron carbonyl concentration should be about 0.1–0.4 ppm. After each week of testing the coupons were recovered from the coupon holder, weighed, and shipped back to Siemens for analysis.

42 ENGINEERING↗

Letter Report: LAW Simulant Development for Cast Stone Screening Test

More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the wastes and immobilize them in a glass waste form. The WTP includes a pretreatment facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the nonradioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at an offsite federal repository. At least a portion (~35%) of the LAW will be converted to glass in the LAW vitrification facility and will be disposed of onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize the wastes destined for each facility. However, a second facility will be needed for the expected volume of additional LAW requiring immobilization. A cementitious waste form known as Cast Stone is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with waste acceptance criteria for the IDF disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. Further, the waste form must be tested to provide the technical basis for understanding the long term performance of the waste form in the IDF disposal environment. These waste form performance data are needed to support risk assessment and performance assessment (PA) analyses of the long-term environmental impact of the waste disposal in the IDF. A testing program was developed in fiscal year (FY) 2012 describing in some detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW (Westsik et al. 2012). Included within Westsik et al. (2012) is a section on the near-term needs to address Tri-Party Agreement Milestone M-062-40ZZ. The objectives of the testing program to be conducted in FY 2013 and FY 2014 are to: • Determine an acceptable formulation for the LAW Cast Stone waste form. • Evaluate sources of dry materials for preparing the LAW Cast Stone. • Demonstrate the robustness of the Cast Stone waste form for a range of LAW compositions. • Demonstrate the robustness of the formulation for variability in the Cast Stone process. • Provide Cast Stone contaminant release data for PA and risk assessment evaluations. The first step in determining an acceptable formulation for the LAW Cast Stone waste form is to conduct screening tests to examine expected ranges in pretreated LAW composition, waste stream concentrations, dry-materials sources, and mix ratios of waste feed to dry blend. A statistically designed test matrix will be used to evaluate the effects of these key parameters on the properties of the Cast Stone as it is initially prepared and after curing. The second phase of testing will focus on selection of a baseline Cast Stone formulation for LAW and demonstrating that Cast Stone can meet expected waste form requirements for disposal in the IDF. It is expected that this testing will use the results of the screening tests to define a smaller suite of tests to refine the composition of the baseline Cast Stone formulation (e.g. waste concentration, water to dry mix ratio, waste loading).

Russell, Renee L.↗

Ranchero Armature Test LA-43-CT-2: PBXN-110 Explosive with no smoothing layer. Lower Slobovia, 6/13/13

This test, documented as LA-43-CT-2, was performed to verify calculations which indicate that the smoothing layer employed for Ranchero Armatures is not required when the slapper point spacing is 18 mm (“On the efficacy of an acrylic layer in the Ranchero armature,” Bob Watt, published in “Ranchero Status Report” LA-14463). [Naming this test –CT-2 acknowledges the previous test performed during 1/11 as the initial test in the series, and we will refer to that test as LA-43-CT-1 in the future.] In addition, higher velocity is achieved when the volume occupied by the smoother is replaced by HE. Two framing cameras were used on the shot, and a 20 point PDV array was fielded to ascertain armature performance in specific conditions. The conclusions are that the armature is smooth enough, and the terminal velocity is increased from ~3.1 mm/μs to ~3.3 mm/μs. The armature tested had casting voids in the HE that were observed in pre-test inspections, and were expected to cause ruptures. (See Appendix 3 of the Ranchero Status Report.) Such “blowouts” were observed in all three locations identified in pre-shot inspection. In addition, ruptures occurred in locations having no pre-shot indication of a problem. These unpredicted blowouts have a difference in appearance from those typically seen. Further visual examination of the inspection radiographs do not reveal the cause, and that is a serious concern for future efforts with such castings. A path forward will include attempting to digitize the inspection x-rays for the charge tested, to see if the cause of the “new” ruptures can be found via computer enhancement. In addition, there are two additional armatures from the same casting lot, and these might also be tested to see if there is some problem with the HE or the armature material in this batch. For now, this plan is on hold, pending the outcome of tests and cost estimates related to using PBX-9501 in place of the PBXN-110. Casting issues would be eliminated in such configurations.

42 ENGINEERING↗

Alternate Reductant Cold Cap Evaluation Furnace Phase II Testing

Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further research and development of this flowsheet eliminated the formic acid, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric-glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12 th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the CEF cold cap and vapor space data to the benchmark melter flammability models; Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: Define new cold cap reactions and global kinetic parameters in support of the melter flammability model development; Quantify off-gas surging potential of the feed; Characterize off-gas condensate for complete organic and inorganic carbon species. After charging the CEF with cullet from Phase I CEF testing, the melter was slurry-fed with glycolic flowsheet based SB6-Frit 418 melter feed at 36% waste loading and was operated continuously for 25 days. Process data was collected throughout testing and included melter operation parameters and off-gas chemistry. In order to generate off-gas data in support of the flammability model development for the nitric-glycolic flowsheet, vapor space steady state testing in the range of ~300-750°C was conducted under the following conditions, (i) 100% (nominal and excess antifoam levels) and 125% stoichiometry feed and (ii) with and without argon bubbling. Adjustments to feed rate, heater outputs and purge air flow were necessary in order to achieve vapor space temperatures in this range. Surge testing was also completed under nominal conditions for four days with argon bubbling and one day without argon bubbling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

FY14 Status Report: CIRFT Testing Results on High Burnup UNF

The objective of this project is to perform a systematic study of SNF/UNF (spent nuclear fuel/or used nuclear fuel) integrity under simulated transportation environments by using hot cell testing technology developed recently at Oak Ridge National Laboratory (ORNL), CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester). Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmarking tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. With support from the US Department of Energy and the NRC, CIRFT testing has been continued. The CIRFT testing was conducted on three HBR rods (R3, R4, and R5), with two specimens failed and one specimen un-failed. The total number of cycles in the test of un-failed specimens went over 2.23 107; the test was stopped as because the specimen did not show any sign of failure. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of used fuel rods in terms of both the curvature amplitude and the maximum of absolute of curvature extremes. The latter is significant because the maxima of extremes signify the maximum of tensile stress of the outer fiber of the bending rod. So far, a large variety of hydrogen contents has been covered in the CIRFT testing on HBR rods. It has been shown that the load amplitude is the dominant factor that controls the lifetime of bending rods, but the hydrogen content also has an important effect on the lifetime attained, according to the load range tested.

fuel vibration integrity↗

Cost Estimate for a Proposed GDF Suez LNG Testing Program

At the request of GDF Suez, a Rough Order of Magnitude (ROM) cost estimate was prepared for the design, construction, testing, and data analysis for an experimental series of large-scale (Liquefied Natural Gas) LNG spills on land and water that would result in the largest pool fires and vapor dispersion events ever conducted. Due to the expected cost of this large, multi-year program, the authors utilized Sandia's structured cost estimating methodology. This methodology insures that the efforts identified can be performed for the cost proposed at a plus or minus 30 percent confidence. The scale of the LNG spill, fire, and vapor dispersion tests proposed by GDF could produce hazard distances and testing safety issues that need to be fully explored. Based on our evaluations, Sandia can utilize much of our existing fire testing infrastructure for the large fire tests and some small dispersion tests (with some modifications) in Albuquerque, but we propose to develop a new dispersion testing site at our remote test area in Nevada because of the large hazard distances. While this might impact some testing logistics, the safety aspects warrant this approach. In addition, we have included a proposal to study cryogenic liquid spills on water and subsequent vaporization in the presence of waves. Sandia is working with DOE on applications that provide infrastructure pertinent to wave production. We present an approach to conduct repeatable wave/spill interaction testing that could utilize such infrastructure.

03 NATURAL GAS↗

Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO 2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO 2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys, hence promoting FCCI between the fuel-clad systems. The other factor was to develop a test bed where multiple candidate alloys could be evaluated within a single irradiation test train, thereby reducing overall costs and increasing efficiency in alloy screening efforts. A collaboration between ORNL and INL was developed to facilitate the completion of the test bed for FCCI testing. The report highlights the activities related to the development of the ATF-1 ORNL FCCI rodlets for irradiation in INL’s ATR as part of the on-going ATF-1 experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Summary Report of Comprehensive Laboratory Testing to Establish the Effectiveness of Proposed Treatment Methods for Unremediated and Remediated Nitrate Salt Waste Streams

The inadvertent creation of transuranic waste carrying hazardous waste codes D001 and D002 requires the treatment of the material to eliminate the hazardous characteristics and allow its eventual shipment and disposal at the Waste Isolation Pilot Plant (WIPP). This report documents the effectiveness of two treatment methods proposed to stabilize both the unremediated and remediated nitrate salt waste streams (UNS and RNS, respectively) at Los Alamos National Laboratory (LANL). The two technologies include the addition of zeolite (with and without the addition of water as a processing aid) and cementation. Surrogates were developed to evaluate both the solid and liquid fractions expected from parent waste containers, and both the solid and liquid fractions were tested. Both technologies are shown to be effective at eliminating the characteristic of ignitability (D001), and the addition of zeolite was determined to be effective at eliminating corrosivity (D002), with the preferred option1 of adding zeolite currently planned for implementation at LANL’s Waste Characterization, Reduction, and Repackaging Facility (WCRRF). The course of this work verified the need to evaluate and demonstrate the effectiveness of the proposed remedy for debris material, if required. The evaluation determined that WypAlls, cheesecloth, and Celotex absorbed with saturated nitrate salt solutions exhibit the ignitability characteristic (all other expected debris is not classified as ignitable). Finally, liquid surrogates containing saturated nitrate salts did not exhibit the characteristic of ignitability in their pure form (those neutralized with Kolorsafe and mixed with sWheat did exhibit D001). Sensitivity testing and an analysis were conducted to evaluate the waste form for reactivity. Tests included subjecting surrogate material to mechanical impact, friction, electrostatic discharge and thermal insults. The testing confirmed that the waste does not exhibit the characteristic of reactivity (D003). Follow-on testing was conducted to demonstrate the effectiveness of zeolite stabilization for ignitable WypAll and cheesecloth debris and additional nitrate salt solutions (those exhibiting the oxidizer characteristic) to demonstrate the effectiveness of the remedy. Follow-on testing also included testing of surrogate materials containing Waste Lock 770, which is present in four of the RNS containers, and potential items of debris such as plywood and Celotex material. Testing to evaluate the effectiveness of the remedy was performed using the specific remediation processes that are planned for use at the WCRRF. Finally, testing was also performed to evaluate the holding capacity of zeolite using a highly acidic surrogate solution and to characterize the composition of gases generated during mixing of zeolite with surrogate solutions. All these tests demonstrated the effectiveness of adding zeolite as the planned remedy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DESIGN AND TESTING OF A DIGITAL REGULATOR FOR FERMILAB MAGNET POWER SYSTEMS

In this thesis, the design of a digitally controlled DC power system for testing conventional and superconducting magnets is proposed. The designed PID controller performances have been tested by the 30kA test stand for superconducting magnets, Vertical Magnet Test Facility (VMTF), which is hosted at the Fermilab Magnet Test Facility (MTF). The system is implemented on a National Instruments CompactRIO and both real-time and FPGA targets are programmed. A full 24-bit PID algorithm is coded and successfully tested by a manual tuning approach. An automated tuning algorithm is then introduced. As it will be shown by simulation and experimental results, the proposed system meets all design specifications. The current loop stability is up to 14 times better than the existing regulator and a control accuracy less than 4 ppm is achieved. Shorted-bus tests of the PID regulator have been successfully performed on the VMTF power system. In order to test the generalization capability of the designed system towards different types of magnets, the system has been easily adapted to and tested on the 10kA conventional magnet test stand (Stand C at Fermilab). As shown by experimental results, the designed PID controller features really high performancesin terms of steady-state accuracy and effectiveness of the tuning algorithm.

43 PARTICLE ACCELERATORS↗

Mechanical Behavior of UO 2 at Sub-grain Length Scales: Quantification of Elastic, Plastic and Creep Properties via Microscale Testing

Techniques were developed to measure properties at sub-grain scales using depleted Uranium Oxide (d-UO2) samples heat-treated to obtain different grain sizes and oxygen stoichiometries, through three main tasks: 1) sample processing and characterization, 2) microscale and conventional testing and 3) modeling. Grain size and crystallography were characterized using Scanning Electron Microscopy (SEM), in conjunction with Electron Backscattering Diffraction (EBSD) and Electron Channeling Contrast Imaging (ECCI). Grains were then carefully selected based on their crystallographic orientations to perform ex-situ micromechanical tests with samples machined via Focused Ion Beam (FIB), with emphasis on micro-cantilever bending. These experiments were performed under controlled atmospheres, to insure stoichiometry control, at temperatures up to 700 °C and allowed measurements involving elastic (effective Young’s modulus), plastic (critical resolved shear stresses) and creep (creep strain rates) behavior. Conventional compression experiments were performed simultaneously to compare with the ex-situ measurements and study potential size effects. Modeling was implemented using anisotropic elasticity and inelastic constitutive relations for plasticity and creep based on kinematics and kinetics of dislocation glide that account for the effects of crystal orientation, and stress. The models will be calibrated and validated using the experimental data. This project provided insight on correlations among stoichiometry, crystallography and mechanical behavior in advanced oxide fuels, provided valuable experimental data to validate and calibrate mesoscale fuel performance codes and also a framework to measure sub-grain scale mechanical properties that should be suitable for use with irradiated samples due to small volumes required. The goals and metrics of the ongoing study of thermo-mechanical behavior in depleted uranium dioxide (d-UO 2 ) outlined in this project have been concluded successfully, resulting in: 1) the successful fabrication, processing, and characterization of large-grained samples with various orientations (up to and including single crystals) having stoichiometric and hyper-stoichiometric O/U ratios; 2) formulation, calibration, and validation of a crystal plasticity constitutive model to describe the creep deformation of UO 2 at the sub-grain length scale (single crystal level) at intermediate temperatures; 3) the successful calibration of a crystal plasticity constitutive model to describe the elasto-plastic deformation of microcantilever beams, also at moderate temperatures. Samples were prepared from natural uranium oxide powder of production-quality provided by Areva. The powder was pressed in a die to a pressure of 100 MPa to produce green pellets with no sintering aids, lubricants, or any other additives. The green pellets were then heated up to 1700 °C under ultra-high purity argon atmosphere (~1 ppm O2). The atmosphere was then changed to 79% Argon, 21% O 2 and the temperature was held at 1700 °C for 2 hours to sinter the pellets under oxidative conditions [1] that are known to increase grain growth kinetics in UO 2 [2]. Samples were then cooled down under Ar-4%H2 atmosphere to reduce the samples back to stoichiometric UO 2 . For macro-scale procedures, testing of UO 2 samples with large grains was performed at 1200 °C using a modified load frame capable of applying dead-weight loads to ensure constant stress conditions, while displacement of the sample produced by the applied load was measured with high precision micrometers to obtain strains. Stress steps were used during testing and the strains were monitored to measured creep strain rates under steady state for each level of stress used, so that stress exponents could be obtained. The results of the mechanical testing, along with sample geometry and crystal orientation of the grains in the samples, as well as post-test sample characterization were used to formulate a viscoplastic model to account for steady state (stage II) creep behavior, along with basic assumptions from crystal plasticity and kinematic constraints due to testing fixtures. In the micro-scale, testing of microcantilever beams at temperatures ranging from 25 to 570 °C was performed in-situ with a scanning electron microscope with a special attachment to apply load and measure displacement while the samples were at temperature. The load-displacement curves obtained showed linear behavior before fracture for all temperatures attempted except 570 °C, where clear deviations from non-linearity were observed before fracture. These deviations were consistently observed for all samples tested for a given orientation. A viscoplastic model was used to account for the presence of inelastic strain, along with basic assumptions from crystal plasticity and beam theory. These models were kept as simple as possible, and results from tests performed in a set of samples with a given crystal orientation were used to calibrate the material constants for the model, while results from a different sample set were then used for validation, thus satisfying the conditions of all main tasks within the parameters of this project. Details of these efforts are outlined in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Pipe Overpack Container Fire Testing: Phase II-A

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible Transuranic (TRU) waste at Department of Energy (DOE) sites. At the request of DOE's Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), SNL started conducting a new series of fire tests in 2015 to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to overpressurize and release its inner content. In 2016, Phase II tests showed that POCs tested in a pool fire failed within 3 minutes of ignition with the POC lid ejecting. These POC lids were fitted with an all-metal (NUCFIL019DS) filter and revealed that this specific filter did not relieve sufficient pressure to prevent lid ejection. For the test phase discussed in this report, Phase II-A, the POCs are exposed to a 30-minute pool fire, with similar configurations to those tested in Phase II, except that the POC lids are fitted with a hybrid metal-polyethylene (UT9424S) filter instead. This report will: describe the various tests conducted in Phase II-A, present results from these tests, and discuss implications for the POCs based on the test results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Simulation of QUENCH-15 and Preliminary Pre-Test Predictions for QUENCH-19

Iron Chromium Aluminum alloys (FeCrAl) have been developed at Oak Ridge National Laboratory as an advanced fuel cladding for light water reactors (LWR). In contrast to zirconium-based alloys employed today, FeCrAl has slower oxidation kinetics and higher strength. A test (QUENCH-19) at Karlsruhe Institute of Technology’s QUENCH facility is planned to test and demonstrate the performance of FeCrAl cladding under postulated accident conditions. This test will replicate the previous QUENCH-15 test but will use FeCrAl instead of zirconium-based materials. During these tests, a small bundle of fuel rod simulators is electrically heated in a steam and argon environment. The temperature history, oxidation thickness, and in-situ hydrogen generation are the primary measurements of interest.The CORA/BWR code was previously developed and utilized in the 1980’s and 1990’s to simulate BWR bundle degradation during the CORA test series. A modified version of the CORA/BWR code, named CORA/BWR-QUENCH, has been developed to model the QUENCH test facility. The CORA/BWR-QUENCH code was used to model the QUENCH-15 test which used zirconium-based materials. The code predictions and test data were found to be in good agreement. The code was then used to model the planned QUENCH-19 test utilizing FeCrAl cladding, shroud, and spacer grids. The simulation predicts the peak cladding temperature to be 1620 K (~1347°C) for QUENCH-19 which is significantly lower than that during QUENCH-15 with the zirconium-based alloy. The predicted hydrogen generation is also significantly lower for the FeCrAl clad QUENCH-19 as compared to QUENCH-15: ~1.7 g versus ~48.8 g. These results align with expectations given the much slower oxidation kinetics of FeCrAl alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Test Plan for the Horizontal Dry Cask Simulator

The thermal performance of commercial spent nuclear fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission(NRC). Canistered dry storage cask systems rely on ventilation between the inner canister and the overpack to convect heat away from the canister to the surrounding environment for both horizontal and vertical configurations. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of the investigation described in this test plan is to produce data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator(HDCS) has been designed to generate this benchmark data and add to the existing knowledgebase. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure(MAWP)rating of 2,400 kPa at400 °C. An existing electrically heated but otherwise prototypic boiling water reactor(BWR), Incoloy-clad test assembly will be deployed inside of a representative storage basket and canister. An insulated sheet metal enclosure will be used to mimic the thermal properties of the concrete vault enclosure used in a modern horizontal storage system. Radial and axial temperature profiles along with induced cooling air flow will be measured for a wide range of decay powers and representative(and higher)cask pressures using various backfills of helium, argon, or air. The single assembly geometry with well-controlled boundary conditions simplifies computational requirements while preserving relevant physics. The proposed test apparatus integrates all the underlying thermal-hydraulics important to defining the performance of a modern horizontal storage system. These include combined-mode heat transfer from the electrically-heated assembly to the canister walls and the primarily natural-convective heat transfer from the canister to the cooling air flow passing through the horizontal vault enclosure. The objective of the HDCS is not to reproduce the performance of a commercial dry storage system for any given set of operational parameters. Rather ,the objective is to capture the dominant physics in a well-characterized test apparatus. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate is of particular importance. While incorporating the best available information based on thermal-hydraulic scaling arguments as well as previous vertical testing, this test plan is subject to changes due to improved understanding or from as built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Battery Testing FY 2017 Annual Progress Report

Abuse tests are designed to determine the safe operating limits of HEV\PHEV energy storage devices. Testing is intended to achieve certain worst-case scenarios to yield quantitative data on cell\module\pack response, allowing for failure mode determination and guiding developers toward improved materials and designs. Standard abuse tests with defined start and end conditions are performed on all devices to provide comparison between technologies. New tests and protocols are developed and evaluated to more closely simulate real-world failure conditions. When scaling from cell to the battery level, a detailed understanding of cell interactions provides insight on safety performance. Single point failures from a cell or group of cells can be initiated by a number of triggers including internal short circuit, misuse or abuse, or component failure at the battery or system level. Propagation of a single failure event (regardless of the initiation trigger) through an entire battery, system, or vehicle is an unacceptable outcome with regards to EV battery safety. In this FY, our work has focused on evaluating the propagation of a single cell thermal runaway event through a battery using a variety of design considerations with an emphasis on passive thermal management impacts. This has been coupled with thermal modeling by NREL for these testing conditions. In addition, alternative failure initiation methods have been evaluated to provide direct comparisons of possible energy injection between modes. This data was compiled to better identify what propagation test method is appropriate given certain battery designs. Expanding the analysis of short circuit current during failure propagation has been done for EV relevant chemistries. Ongoing test development and validation to obtain these values has been achieved. While robust mechanical models for vehicles and vehicle components exist, there is a gap for mechanical modeling of EV batteries. The challenge with developing a mechanical model for a battery is the heterogeneous nature of the materials and components (polymers, metals, metal oxides, liquids). Our work will provide empirical data on the mechanical behavior of batteries under compressive load to understand how a battery may behave in a vehicle crash scenario. This work is performed in collaboration with the U.S. Council for Automotive Research (USCAR) and Computer Aided Engineering of Batteries (CAEBAT). These programs have supported the design and development of a drop tower testing apparatus to close the gap between cell/string level testing and full scale crash testing with true dynamic rate effects.

25 ENERGY STORAGE↗

Pipe Overpack Container Fire Testing: Phase II-A

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible Transuranic (TRU) waste at Department of Energy (DOE) sites. At the request of DOE's Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), SNL started conducting a new series of fire tests in 2015 to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to overpressurize and release its inner content. In 2016, Phase II tests showed that POCs tested in a pool fire failed within 3 minutes of ignition with the POC lid ejecting. These POC lids were fitted with an all-metal (NUCFIL019DS) filter and revealed that this specific filter did not relieve sufficient pressure to prevent lid ejection. For the test phase discussed in this report, Phase II-A, the POCs are exposed to a 30-minute pool fire, with similar configurations to those tested in Phase II, except that the POC lids are fitted with a hybrid metal-polyethylene (UT9424S) filter instead. This report will: describe the various tests conducted in Phase II-A, present results from these tests, and discuss implications for the POCs based on the test results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Production Facility Prototype Blower 1000 Hour Test Results IV

Long duration operations tests of the Aerzen GM 12.4 roots style blower in a closed loop configuration provides valuable data and lessons learned for long-term operation at the Mo-99 production facility. The blower was operated in a closed loop configuration with the flow conditions anticipated in plant operation with a Mo-100 target inline. The additional thermal energy generated from beam heating of the Mo-100 disks was not included in these tests. Twelve long duration 500 to 1000 hour tests have been completed since the first test was performed in January of 2016. All twelve long duration tests have proven successful in exposing preventable issues related to oil and helium leaks. All blower tests to this date have resulted in stable blower performance and consistency. A summary of the results for the end of the ninth test to the twelfth test is included in this report. Complete details for the previous tests are shown in the reports, Production Facility Prototype Blower 1000 Hour Test Results, II, and III.

42 ENGINEERING↗

FY19 Status Report on Creep Test Data on Commercial Heat of Alloy 790

This report provides an update on the status of the creep rupture testing on the first commercial heat of Alloy 709 in support of the Alloy 709 Code Case development. The report is a Level 3 deliverable in FY19 (M3AT-19AN020502041), under the Work Package AT19AN0205204, “A709 Code Case Testing” performed by Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program. A creep test laboratory has been set up at ANL that comprises of 15 indirectly loaded test frames capable of testing ASTM standard-size specimens of the commercial heat of Alloy 709. The systems are equipped with high temperature furnaces capable of operation up to 1000°C in an air environment. In FY19, 15 creep-rupture tests of the ESR-1100°C solution annealed material, requiring the use of Type K thermocouples, were initiated to support the 100,000-hour Code Case. At the time of writing this report, eight tests were completed. This report presents the data from those completed tests and also microstructural characterization of the fractured specimens. During FY19, efforts were being made to add Type S thermocouples to five of the creep test frames, for performing longer term tests at higher temperatures.

36 MATERIALS SCIENCE↗