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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Supercritical carbon dioxide tubular flow under temporally varying thermal boundary condition

During transient operation of fusion power plants the amount of thermal energy transferred from plasma to surrounding blanket modules will be varied over time, and will affect behavior of the working fluid inside the blanket and power conversion system where the coolant is in a supercritical state. Transient behavior of the power is in pulsed state in tokamak. The Optimized Supercritical Cycle Operation (OSCO) loop is constructed to investigate the thermohydraulic characteristics of the supercritical fluid under temporally varying thermal boundary condition. In this study the tube outer wall temperature data are measured for abrupt change in thermal power as a preliminary power transient test. The OSCO test conditions are selected to include the erratic behavior of the supercritical fluid under pseudo-critical condition during transient. In order to incorporate the delayed response of utilized thermocouples, a time constant is applied to adjust the obtained results. Along with the experimental study, computational fluid dynamic software is used to perform detailed analysis over the test section geometry. The preliminary test results are presented for comparison against the available correlations from the literature. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Baseline Assessment of TREAT for Modeling and Analysis Needs

TREAT is an air-cooled, graphite moderated, thermal, heterogeneous test facility designed to evaluate reactor fuels and structural materials under conditions simulating various types of nuclear excursions and transient undercooling situations that could occur in a nuclear reactor. After 21 years in a standby mode, TREAT is being re-activated to revive transient testing capabilities. Given the time elapsed and the concurrent loss of operating experience, current generation and advanced computational methods are being applied to begin TREAT modeling and simulation prior to renewed at-power operations. Such methods have limited value in predicting the behavior of TREAT without proper validation. Hence, the U.S. DOE has developed a number of programs to support development of benchmarks for both critical and transient operations. Extensive effort has been expended at INL to collect detailed descriptions, drawings and specifications for all aspects of TREAT, and to resolve conflicting data found through this process. This report provides a collection of these data, with updated figures that are significantly more readable than historic drawings and illustrations, compositions, and dimensions based on the best available sources. This document is not nor should it be considered to be a benchmark report. Rather, it is intended to provide one-stop shopping, to the extent possible, for other work that seeks to prepare detailed, accurate models of the core and its components. Given the nature of the variety of historic documents available and the loss of institutional memory, the only completely accurate database of TREAT data is TREAT itself. Unfortunately, disassembly of TREAT for inspection, assay, and measurement is highly unlikely. Hence the data provided herein is intended serve as a best-estimate substitute.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Phased Development of Accident Tolerant Fue

The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modeling predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology readiness levels (TRL) defined for nuclear fuel development. TRL 1–3 corresponds to the “proof-of-concept” stage (Phase 1), TRL 4–6 to “proof-of-principle” (Phase 2), and TRL 7–9 to “proof-of-performance” (Phase 3). This paper will provide an overview of the anticipated activities within each phase of development and will provide an update on the current ATF development status.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Phased development of accident tolerant fuel

The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modeling predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology readiness levels (TRL) defined for nuclear fuel development. TRL 1-3 corresponds to the 'proof-of-concept' stage (Phase 1), TRL 4-6 to 'proof-of-principle' (Phase 2), and TRL 7-9 to 'proof-of-performance' (Phase 3). This paper will provide an overview of the anticipated activities within each phase of development and will provide an update on the current ATF development status. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A simulation of a steam generator tube rupture in AP1000 nuclear power plant based on RELAP5-MV

In the operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, steam generator tube rupture (SGTR) is one of the main accidents. In this research, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SGTR accident in AP1000 nuclear power plant (NPP) is analyzed by Relap5-MV thermo-hydraulics code. The transient test scenario is modeled using the RELAP5-MV for the AP1000 NPP. The calculation was performed up to a transient time of 3,000.0 s. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially beginners. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Detailed characterization of a PWR fuel rod at high burnup in support of LOCA testing

Experimental investigations of the fuel microstructure and volatile fission products along the pellet of a high burnup specimen (local burnup 76 GWd/tHM) have been conducted to support future transient testing. Detailed microscopy examinations have been carried out at different length scales. Transmission electron microscopy has highlighted a significant amount of damage across the entire radius with the formation of networks of dislocations. The optical and scanning electron microscopy determined the formation of three zones in the pellet with different characteristics. An intermediate region with no grain subdivision, lower porosity than the central zone porosity and high retained fission gas in nanometric bubbles and in the matrix was present between r/r0 ˜ 0.55 and 0.8. The high retention of gas in this region might suggest that the region will be prone to fine fragmentation, in addition to the HBS. An abrupt transition in the structure was observed at mid radius, with a third region developing from the mid radius to the pellet center. In this part of the pellet, metallic and grey phases with size between hundreds of nanometers and a few micrometers have formed at grain boundaries. The majority of fission gas has been released from the grain matrix and the original grains have polygonised, forming sub-grain domains separated by low-angle grain boundaries. No final explanation can be given for the polygonization occurring in the center, but on the basis of the irradiation history and the analysis of all the post irradiation examination (PIE) data, it is postulated that the polygonization within the original grains is an effect of dynamic recovery occurring at high temperature in the fuel center.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Uncollided Flux Techniques for Discrete-Ordinate Radiation Transport Solutions in Rattlesnake

One of the only real-time-resolved measurement tools used at the Transient Test Reactor (TREAT) is the fast-neutron hodoscope. The hodoscope was used for monitoring and measuring fuel motion during a transient pulse. The hodoscope is a line of sight detection and imaging system that provides both temporal and spatial resolution of fuel motion during transients, and in-place measurement of fuel distribution during and after transient experiments. However, the hodoscope relies on fast neutron streaming out of the reactor core, which provides a challenge to transient modeling and simulation. However, use of a first collision source approach can be used to overcome this shortcoming. Hence, the TREAT modeling and simulation team has initiated research to implement such capabilities in the neutron transport code Rattlesnake. This report reviews uncollided flux techniques (first and last collision methods) to be implemented in the Rattlesnake SN code in order to mitigate ray effects in modeling the TREAT reactor+hodoscope system. Angular discretization techniques (SN and PN) for the transport equation are notoriously poor at capturing effectively streaming effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

BENEFIT with Northeastern University: HVAC Hardware-in-the-Loop Experimental Testing of a Heat Pump and Air Conditioner

This dataset includes HVAC Hardware-in-the-Loop (HIL) experimental results for a single stage, SEER 16, HSPF 9.5, 3-ton single-speed air source heat pump with 15 kW of backup auxiliary heating tested in both cooling and heating mode, and a two stage, SEER 21, 2-ton central air conditioner tested in cooling mode for a set of outdoor temperatures and indoor setpoint temperatures. In addition to these tests, experimental tests focused on the operation of auxiliary heating for the heat pump for winter condition were also conducted. The laboratory experiments for transient testing of the heat pump and air conditioner were conducted using the two HIL systems in the Systems Performance Laboratory (SPL) at NREL’s Energy Systems Integration Facility (ESIF). Further information on laboratory design and capabilities of the SPL along with the architecture of HVAC HIL system can be found in: Sparn, B. F. 2018. Laboratory Resources and Techniques to Evaluate Smart Home Technology (No. NREL/CP-5500-71696). National Renewable Energy Laboratory (NREL), Golden, CO (United States). https://www.nrel.gov/docs/fy18osti/71696.pdf and the experimental setup and validation of HVAC HIL platform can be found in: Ramaraj, S. and Sparn, B. 2022. Validation of HVAC Hardware-In-the-Loop Simulation for Advanced Control Strategies in Smart Homes (No. NREL/CP-5500-82562). National Renewable Energy Lab (NREL), Golden, CO (United States). https://www.nrel.gov/docs/fy22osti/82562.pdf. These experimental results can be used to validate how we currently model the cycling behavior of heat pumps and air conditioners. Additionally, many demand response programs implement heat pump and air conditioner control by changing the thermostat set point – these data may also be used to verify our models for heat pump and air conditioner demand response control are implemented correctly. The Test_Matrix file describes all the indoor and outdoor test conditions for heat pump and air conditioner and the file names of data sets include information about the test conditions. A wide range of outdoor air temperatures were chosen to accommodate summer and winter conditions. In addition to operating the HVAC equipment with different outdoor temperatures, we also operate the system with different indoor temperature set points to represent different grid signals or different operating conditions. For cooling conditions, the baseline set point is 72°F. To represent Load Up signals, the setpoint is changed to 68°F. The Load Shed set point is 76°F. For heating conditions, the baseline set point was assumed to be 68°F. The Load add set point is 72°F and the Load shed set point is 64°F. The starting indoor temperature for cooling conditions was set ~2°F above the indoor setpoint temperature so that the equipment turned on quickly. Similarly, the initial indoor temperature was set ~2°F lower than setpoint for heating mode tests to ensure that heating began quickly. The return air temperature was assumed to be equal to the indoor setpoint temperature in all cases. The experimental data are sampled at 1-second intervals. The data from ecobee thermostat at 5-minute interval are resampled and added to the corresponding file. The content of each data set is as follows: • T_Return (C): Measured return air temperature [C] • T_Return_SP (C): Return air temperature setpoint from E+ model, sent to HIL [C] • T_Supply (C): Measured supply air temperature at evaporator outlet [C] • T_Outdoor (C): Measured outdoor air temperature [C] • T_Outdoor_SP (C): Outdoor air temperature setpoint from weather file, sent to HIL [C] • T_Indoor (C): Measured indoor air temperature [C] • T_Indoor_SP (C): Indoor air temperature setpoint from E+ model, sent to HIL [C] • Outdoor Unit Power (W): Measured power of the outdoor unit [W] • Indoor Unit Power (W): Measured power of the indoor unit [W] • Evaporator Airflow Rate (CFM): Measured evaporator or indoor unit airflow rate sent to E+ model [CFM] • Cooling/Heating Capacity (kW): Calculated cooling/heating capacity sent to E+ model [kW] • T_SP_Thermostat (C): Thermostat cooling/heating setpoint temperature [C] • T_Indoor_Thermostat (C): Thermostat indoor air temperature [C]

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2018

Integrated Boosting and Hybridization for Extreme Fuel Economy and Downsizing

The objective of this project was to develop, demonstrate, and evaluate commercialization of a highly efficient downsized engine by electrification of the air delivery and waste heat recovery system and optimizing energy usage to achieve significant fuel economy improvement at a commercially viable cost. The system consists of an Electrically Assisted Variable speed Supercharger (EAVS), driven by an electric motor and the engine through a planetary gear train, that provides engine boost pressure and provides the mild hybrid functions of engine start/stop, torque assist, and brake energy recovery. In the exhaust path of the engine is the electric Waste Heat Recovery (eWHR) system, a roots-based expander driven by another electric motor, that enables exhaust gas energy recovery along with controlled back pressure on the engine. The two electric motors are powered by a 48V rechargeable lithium-ion battery. Further, the EAVS system coupled to an EGR valve enables very precise high exhaust recirculation rates to control the NOx emissions. Through two budget periods, the project team of Eaton, Isuzu, Southwest Research Institute and AVL designed and developed the EAVS and eWHR systems, integrated with an Isuzu diesel engine, and validated the target fuel economy performance and emissions by engine dynamometer testing and vehicle level simulations. GT Power model-based analyses results were used to size the EAVS and eWHR systems for the 1.9L Isuzu Spark engine. The two systems were built and the functional testing was carried out to verify the boosting performance of the superchargers. Following the baseline engine dynamometer testing, the two systems were integrated with the engine and the fuel economy performance and emissions were determined by combined steady state and transient testing over three drive cycles, namely, FTP75, NEDC and HWFET. The EAVS/eWHR system demonstrated 4% improvement in fuel economy for the NEDC cycle in engine dyno testing, correlating well with vehicle simulation results (5.3%). The simulation results indicated that the combination of hybrid functions and drivetrain optimization, fuel economy improvement of more than 20% was feasible with the system. The commercialization study to evaluate the cost of the system and the scalability for different engine/vehicle classes was planned for the final budget period but not completed as the project was closed at the end of budget period two.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2021

High-burnup fuel stress analysis prior to and during a LOCA transient

High-burnup fuel fragmentation and pulverization have been experimentally observed during simulated loss-of-coolant accident (LOCA) transient testing of high-burnup (>62 GWd/tU) nuclear fuel. The cause of this phenomenon is generally accepted to be the formation of stresses arising from the off-normal conditions. However, the primary mechanism(s) contributing to the stress formation is/are still being debated. Leading hypotheses suggest that the concurrence of fission gas bubbles and high burnup structure on the pellet periphery (i.e., the rim) leave the fuel susceptible to pulverization. Other hypotheses suggest that pulverization is artificially inflated during externally heated tests that do not reproduce prototypic fuel stresses and radial temperature profiles as the fuel transitions from full-power nuclear heating to decay heating. This work performs finite element analyses of fuel pellet temperatures and macroscopic stresses prior to and during a LOCA transient to identify (1) operating conditions that may exacerbate the fuel pellet stress state, (2) differences in predicted stresses with nuclear vs. furnace heating, and (3) how fuel cracking impacts the subsequent stress state in the pellet fragments. Analyses predict that macroscopic transient fuel stresses resulting from constrained thermal expansion will remain below the fracture stress of fresh UO 2 once the fragment size reaches ~4 mm or less. Calculated fuel stresses from simulated LOCA tests performed with representative nuclear vs. electrical heating rates are shown to be equal within 5% and essentially unchanged for all times after the first 27 seconds of the transients. Because the fuel stresses do not change once the temperature profile becomes uniform, macroscopic transient stresses would be no different than those following a normal shutdown. Because pulverization of high burnup fuel does not occur prior to a LOCA test, pulverization of high burnup UO 2 cannot be initiated by differential thermal stresses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Flow boiling transient critical heat flux tests with stainless steel and FeCrAl: Transient correlation implementation, model calibration, and sensitivity analysis

In this study, steady-state and transient internal-flow critical heat flux (CHF) experiments were carried out under two flow conditions, at atmospheric pressure, on 316 stainless steel (316-SS) and iron-chromium-aluminum (FeCrAl) tubes. Slow power transients at a low mass flow (300 kg/m 2 - s) with very low subcooling (X e =-0.0054) generated premature CHFs, which were prevented by faster power transients and high mass flow (1000 kg/m 2 - s). The measured transient CHFs increased linearly with increasing power transients, compared with the steady-state CHF. Yet, the wall superheat at the CHFs decreased with faster power transients. Transient CHF correlations highlighting heterogeneous spontaneous nucleation were calibrated to the measured CHFs and compared with other existing correlations. Transient CHF multipliers were acquired from pool and flow boiling empirical CHF correlations that were generated. The multipliers were applied to the RELAP5-3D nuclear system code to analyze the discrepancies between the measured data and the predicted CHF and post-CHF behavior, which improved peak cladding temperature predictions by 24.6%. A variance-based global sensitivity study perturbing the experimental uncertainties and cladding material thermal properties showed the diminishing influence of flow boiling heat transfer with increasing power transients, highlighting the significance of the volumetric heat capacity for the cladding integrity during transients. Transient CHF correlations were applied to the most limiting design basis accident: a hot-zero-power reactivity-initiated accident on a generic pressurized water reactor RELAP5-3D model.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT

The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Accelerated testing of module-level power electronics for long-term reliability

This work has applied a suite of long-term-reliability accelerated tests to a variety of module-level power electronics (MLPE) devices (such as microinverters and optimizers) from five different manufacturers. This dataset is one of the first (only the paper by Parker et al. entitled “Dominant factors affecting reliability of alternating current photovoltaic modules,” in Proc. 42nd IEEE Photovoltaic Spec. Conf., 2015, is reported for reliability testing in the literature), as well as the largest, experimental sets in public literature, both in the sample size (five manufacturers including both dc/dc and dc/ac units and 20 units for each test) and the number of experiments (six different experimental test conditions) for MLPE devices. The accelerated stress tests (thermal cycling test per IEC 61215 profile, damp heat test per IEC 61215 profile, and static temperature tests at 100 and 125 °C) were performed under powered and unpowered conditions. The first independent long-term experimental data regarding damp heat and grid transient testing, as well as the longest term (>9 month) testing of MLPE units reported in the literature for thermal cycling and high-temperature operating life, are included in these experiments. Additionally, this work is the first to show in situ power measurements, as well as periodic efficiency measurements over a series of experimental tests, demonstrating whether certain tests result in long-term degradation or immediate catastrophic failures. Lastly, the result of this testing highlights the performance of MLPE units under the application of several accelerated environmental stressors.

14 SOLAR ENERGY↗
DOE OSTI2020

ARCS Replacement at TREAT for TRTR 2020 - Abstract

The Transient REActor Test (TREAT) facility was restarted in 2017 to perform reactor fuels testing under transient conditions, following shutdown in 1994 and 23 years in non-operational standby. TREAT is a transient reactor that, like most pulsing reactors, can create natural shaped power pulses. However, TREAT provides the ability to shape the power pulse and control the energy deposition rate on the test specimens by automatically controlling the amount of reactivity insertion over time. The original automatic reactor control system (ARCS) was created in the 1980s using INTEL 8086 CPU boards and software written in assembly code. The need to maintain the system for the next 40+ years of TREAT life led to the decision to replace ARCS with modern hardware while maintaining the original functionality. After a prototyping effort, the replacement was successfully completed in the first quarter of 2020. The process and results of that effort are reported here.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Scoping Study for Fast Flux Testing in the Advanced Test Reactor

The value of fast spectrum reactors remains prominent in the nuclear technology portfolio. The performance of these reactors can be maximized with advancements in nuclear fuel technologies, but development of these technologies is currently held back by lack of fast spectrum test reactors available to the United States. Spectral modification of experiment positions in the thermal spectrum Advanced Test Reactor (ATR) has long been used to support fast reactor fuel development, but these methods have not been progressed to their full potential. This study investigated the use of concentric rings of aluminum-clad fuel plates in ATR flux traps and thermal neutron absorbing filters to increase fast neutron flux on test specimens. This concept was termed the Boosted Energy Advanced Spectrum Test (BEAST). This approach will enable irradiation of advanced fuel designs in prototypic-length fuel pins and representative flux environment to support post irradiation exams, enable transient testing, and produce the type of data that will permit lead test assembly irradiations in true Sodium Fast Reactors (SFRs) when they become available. Neutronic predictions were performed to investigate BEAST design options and thermal hydraulic models were produced to ensure feasibility of BEAST. Two versions were considered based on the geometric limitations of ATR’s small and large flux traps. The small version was found to be preferable due to slightly higher fast flux and fast-to-thermal neutron ratio. Perhaps more influentially, the small flux trap option was also preferred to avoid conflict with ongoing very high temperature reactor fuel irradiation programs in ATR’s large northeast flux trap. The small flux trap option provided less than half the test volume of the large version, but still had adequate volume for seven SFR pins in cross section which could be stacked two-high in ATR’s 1.2m long core to accommodate up to 14 EBR-II size pins. The preference for the small flux trap configuration should be revisited if additional collaborative test programs emerge with the need to irradiate a significant volume of additional specimens. Calculations were performed regarding a lithium deuteride ring to convert thermal neutrons into 14 MeV fusion neutrons. At the time this report was written these calculations were partially complete and it remains to be seen whether the concept would be worth including in BEAST. Given the preference for the small flux trap option, which does not afford enough room for the 14 MeV ring, it was concluded to defer future work on the lithium deuteride ring. This decision could be revisited if fusion material research programs emerge for collaborative testing in BEAST. A cadmium-lined specimen holder design was found to be adequate in filtering thermal neutrons and preferred over other neutron absorbers based on past experience with cadmium baskets. It was acknowledged that cadmium-bearing hardware would become depleted and need to be replaced occasionally, which appeared feasible from a mechanical design perspective. Neutronic studies investigated different enrichment levels in the booster fuel using uranium-molybdenum alloy dispersion fuel which has performed well in past ATR irradiations. Both options were able to drive fuel pins to SFR-like fission heating rates. The high enriched booster fuel option outperformed the low enriched option by ~20% on key metrics including fast flux and fast-to-thermal ratio, but the low enriched option was favored in order to broaden options for potential fuel suppliers. The preferred BEAST design options including cadmium filter with low enriched booster fuel in the small flux trap configuration was predicted to achieve 6.2E14 n/cm2sec fast flux (>0.1 MeV) with a fast-to-thermal ratio of 44.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

FY23 Status of Quality Assurance Plan for Out-of-Pile Test Data

The DOE Advanced Reactor Technology program has supported recovery and preservation of legacy metallic fuel data collected as part of the US fast reactor program, recognizing it as essential to development and licensing activities for advanced fast reactors. Databases were established as organized collections of experimental records and data generated from in-pile experiments at EBRII, FFTF, and TREAT as well as related out-of-pile examinations of irradiated fuels. The Out-of-Pile Transient Database (OPTD), includes records of over 150 out-of-pile furnace transient tests on metallic fuels conducted at Argonne’s Alpha-Gamma Hot Cell Facility to evaluate their transient performance and characterize fuel/cladding interaction. The database is accessible to registered users from US universities, laboratories, and nuclear industry. Because the data in OPTD has not been formally qualified, its applicability and ability to support licensing activities is limited. This report outlines progress and plans to quality assure data in OPTD, maximizing its impact for model validation and verification as well as qualification of fuels for safe and effective use in advanced reactor designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents

The mechanical response of a nuclear-grade silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite was investigated under mechanical loading conditions mimicking the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA). In a RIA, cladding deformation and failure can be induced by the rapid thermal expansion of the nuclear fuel. A pulse-controlled modified-burst test was used to investigate RIA-like PCMI scenarios on SiC/SiC composite samples at pulse widths from 12 to 100 ms.The strain-driven nature of the cladding sample deformation was due to the rapid internal pressurization and subsequent expansion of a secondary tube. A digital-image correlation technique was used to measure strains from the speckle-painted outer surface of the tubes. The failure strains of samples tested at slower rates, such as RIA event durations of 52 and 100 ms, showed good agreement with the literature-reported values for similar composites tested at slow strain rates. Additionally, the failure strain showed good agreement with reference expansion-due-to-compression tests at slow strain rate. However, a decrease in the failure strain was determined for the fast-rate (12 ms) tests. This indicated that the failure strain of these composites might be influenced by the strain rate during RIA-like events. In conclusion, the failure strains observed in the tests corresponded to local energy depositions of approximately 50 cal/g UO 2 from hot zero power, with an initial condition of pellet–cladding gap closure prior to the event. In-pile transient testing of these concepts that would result in hoop strain due to PCMI in the range of 0.5–1.0% is recommended.

36 MATERIALS SCIENCE↗