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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

AmeriFlux FLUXNET-1F US-xSL NEON North Sterling, CO (STER)

This is the AmeriFlux Management Project (AMP) created FLUXNET-1F version of the carbon flux data for the site US-xSL NEON North Sterling, CO (STER). This is the FLUXNET version of the carbon flux data for the site US-xSL NEON North Sterling, CO (STER) produced by applying the standard ONEFlux (1F) software. Site Description - The North Sterling site, located outside of Sterling in Northeastern Colorado, is characterized by flat terrain and represents an agronomic site within urban and formerly urban areas. The site is at the edge of a non-tilled experimental field that is used for the long-term sustainable Dryland Agroecosystems Project (DAP), which was initiated in 1985 at three sites in eastern Colorado (Sterling, Stratton, and Walsh) to evaluate the effects of cropping intensity on production, water use efficiency and selected soil chemical and physical properties. The DAP site was established in 1985 and was chosen because of representative soils present in the catena.

Network), NEON (National Ecological Observatory↗
DOE OSTI2021

Strengthening Technical Safeguards of Fresh Fuel Through International Cooperation

Over time, nuclear fuel designs have shifted towards having higher initial enrichments and a greater number of burnable poison rods. This enables increased burnup for commercial reasons. These changes have made safeguards measurements of the 235 U content in modern fresh fuel more challenging.This is addressed here through a re-evaluation of the UNCL (Uranium Neutron Collar–Light Water Reactor Fuel)poison rod correction factor. Coefficients of the poison rod correction factor were updated by simulating response of AngraII PWR fuel with a wide range of both burnable poison rods and Gd 2 O 3 contentper rod. Benchmark comparisons are made to experiments performed at Resende (Brazil) and CNEN/LASAL in Brazil. By updating these coefficients, while maintaining the same mathematical form to allow INCC to be used without software modification,it was possible to reduce error in experimental 235 U linear density assay by an order of magnitude.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Submersible Multistage Centrifugal Pump for Versatile Test Reactor Cartridge Test Loop

Submersible multistage centrifugal pumps are ideal for pumping in narrow confined spaces and achieving necessary head pressures and flow rates. Once a diameter is determined then manipulation of the number of stages and motor speed are all that are required to meet desired flow conditions. The Versatile Test Reactor (VTR) closed loop cartridge systems will need forced convection cooling independent of the main reactor. A multistage centrifugal pump can meet the necessary flow rates and pumping pressures while minimizing space taken. The pump considered for this work was based off a deep well submersible pump, a variation of a multistage centrifugal pump. We experimented with two pump sizes, 5 cm (2 inch) and 7.5 cm (3 inch) diameters. These diameters were chosen to fit into the inner diameter of standard 5 and 7.5 (2 and 3 inch) Schedule 40 pipe, respectively. This made the design for the test loop both simpler and less expensive as the need for an engineered pump housing was eliminated. Initial test cartridge planning indicated space for only a 5 cm (2 inch) diameter pump, though early testing of this size showed the need for an abnormally high-speed and high-power motor. Fine tuning of the cartridge design allowed a pump size increase to 7.5 cm (3 inches), which was the pump size most extensively tested in this work. The test loop is composed of various sizes of PVC and aluminum piping components in a loop configuration. The pump is driven by a Pittman 250 W (1/3 horsepower) electric motor with maximum speed of 3,450 RPM. Testing consisted of running the pump at a constant motor speed while varying a control valve to restrict flow through the loop, with differential pressure and flow rate recorded. This was done for one and two stage configurations for the 5 cm (2 inch) diameter impeller design and one, two, and three stage configurations for the 7.5 cm (3 inch diameter) impeller design, respectively. Due to pumping power requirements, two and three stage 7.5 cm (3 inch) diameter impeller testing at higher flowrates lowered the motor speed substantially. In regions where motor speed could not be maintained constant, the data were discarded. The test loop was also reconfigured to allow for the pump to be tested for pressure drop in a stalled or inoperable (0 RPM) flow condition. Demonstration of adequate natural convection cooling of the test cartridge fuel type is necessary under accident conditions, and this will depend upon the flow resistance through the impeller assembly when the pump is not operating. Thus, accurate knowledge of the effective impeller assembly loss coefficient is important for safety evaluations. The test loop was modified to provide water inlet and outlets on either side of the pump impeller stack, and a metered flow of lab water was provided in order to measure the pressure drop across the cartridges as a function of flowrate. Data from the pump head curve testing developed as part of this work and supported by analysis using pump head affinity laws indicates that a three stage 7.5 cm (3 inch) pump impeller design will meet target requirements for coolant flow within the VTR cartridge sodium cartridge at full power conditions [1] with margin; this corresponds to a flowrate of 45 l/min (12 gpm) at a pressure drop of 6.1 m (20 feet) of water head. The results of the pressure loss measurements across the impeller assembly when the pump is stationary (i.e., at 0 RPM) indicate that the pressure loss coefficient is 0.921 for a two impeller stack configuration; this value is calculated based on the flow velocity through the minimum available flow area within a single stage of the impeller which corresponds to 1.4 cm2.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Modeling and Simulation of Xe-100-type Pebble Bed Gas-Cooled Reactor with SCALE

The US Department of Energy (DOE) announced the Advanced Reactor Demonstration Program (ARDP) to accelerate the deployment of advanced reactor concepts. Awardees of ARDP funds are expected to demonstrate the operation of an advanced reactor within 7 years of receiving the award. X-Energy’s advanced reactor concept, the Xe-100, was selected as one of two advanced reactor concepts to receive funding to demonstrate the operation of its high-temperature gas-cooled pebble-bed reactor before the end of this decade. As a result of this push to bring advanced reactors to maturation and commercialization, transition and deployment scenario studies are being performed under the Systems Analysis and Integration (SA&I) campaign within the DOE Office of Nuclear Energy (DOE-NE) to evaluate the transition of the current US commercial fleet of light-water reactors (LWRs) to a future fleet of advanced reactors consisting of a mix of ARDP type reactor concepts and advanced LWRs. To accurately evaluate the front- and back-end resource requirements, it is important to perform reactor physics calculations to determine the discharge burnup and isotopic content, fuel residence time, as well as other parameters. For this purpose, a summer project funded by the SA&I campaign allowed for the setup of SCALE models for full-core Xe-100 type high-temperature gas-cooled pebble-bed reactor and a Xe-100 type slice using publicly available information. The core-averaged equilibrium compositions and zone-wise equilibrium compositions for the slice and 3D models, respectively, were obtained following an iterative depletion method developed by Bostelmann et al. using SCALE’s reactor physics sequence TRITON. The slice model was used with TRITON to generate burnup-dependent cross section libraries at different temperatures which can be used with SCALE’s ORIGAMI code to rapidly determine fuel inventory and therefore to perform quick sensitivity studies on parameters such as the pebble location in the core. The SCALE/TRITON transport and depletion calculation for the Xe-100 type slice model indicates that the isotopic concentrations are in good agreement at 1,300 effective full power days (EFPD) for 235 U. An analysis of 236 U results match 239 Pu results would seem to indicate a typographical error in Mulder and Boyes wherein the reported results of 236 U and 239 Pu are reversed. In addition to SCALE/TRITON calculations, a new capability within SCALE/ORIGAMI for the simulation of pebble-bed reactors was used to study the burnup sensitivity with respect to the pebble pathway through the core. The SCALE/ORIGAMI results show that pebbles that travel closer to the reflector for the entire depletion history have a higher burnup than pebbles that travel through the middle of the core because of the higher thermal to fast flux ratio near the reflector. Consequently, a pebble’s burnup is strongly affected by the pebble’s pathway for each pass. Additional phenomena such as temperature distributions in the core and different travel times of the pebbles in the individual radial zones further affect the burnup distribution. The sensitivity of the discharge vector to the pebble pathways taken during each pass can be evaluated in the future using SCALE/ORIGAMI now that the SCALE inputs have been established.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Review and Summary of Corrosion Behavior for Aluminum-Clad Spent Nuclear Fuel in Dry Storage

A description of the corrosion behavior of aluminum alloys used for cladding on aluminum-based, aluminum-clad nuclear fuel used in research reactors under potential dry storage conditions has been compiled. An evaluation is made of potential additional corrosion with water postulated to not be removed during drying. The relative humidity (RH) produced by typical dryness criteria for industrial spent nuclear fuel drying is expected to be low, particularly at high temperatures (17% RH at 20°C and lower at higher temperatures). Existing corrosion data suggests that negligible vapor-phase corrosion of aluminum is expected at such low humidity, even for nominally bare aluminum surfaces. That is, the predicted relative humidity from free water (e.g., 17% RH) is below the “critical” relative humidity (~40% RH at room temperature), below which no significant corrosion is observed for bare aluminum. Furthermore, reported room-temperature corrosion rates are very low even under saturated (100% RH) water vapor. Existing results showing significant vapor-phase corrosion corresponded specifically to conditions of high temper ed with high relative humidity. A relatively large reservoir of (chemically bound) water exists in the aluminum (oxy)hydroxide films on the SNF cladding surface. The estimated total could saturate the gas even at relatively high temperature if fully released; however, its release as molecular water is expected to only be plausible if the temperature during storage exceeds both the drying temperature and the threshold for thermal decomposition of the trihydroxides (~220°C). Therefore, the combination of high temperature and high relative humidity that could drive significant corrosion is considered implausible during sealed dry storage following an appropriate drying process, to include >220°C drying for canisters that may approach or exceed this temperature during storage. Vapor-phase testing of aluminum samples with an adherent (oxy)hydroxide layer, prepared in liquid water to resemble those on actual aluminum-clad spent nuclear fuel (ASNF), did not observe evidence of additional corrosion even at combined high-temperature (up to 180°C) and high-humidity (up to 100% RH) conditions. Instead, small net mass losses were observed for most specimens and attributed to dehydration of the samples, which had been air-dried only prior to testing. ASNF being moved to dry storage is expected to have an existing (oxy)hydroxide film, suggesting that the cladding will be less prone to additional corrosion than bare aluminum metal. If significant corrosion did occur during sealed storage, the overall extent and impact is expected to be small. Corrosion post-closure of a canister would not alter the total amount of hydrogen in the canister, so it would have no effect on the maximum H 2 release already assessed in existing bounding calculations. In addition, the total amount of water in a 25-µm dense bayerite film would consume a only ~8 µm additional aluminum metal on average, if fully consumed by oxidizing aluminum metal to Al 2 O 3 . These conclusions are consistent with ASNF-in-canister simulations to date, which included a reaction pathway for corrosion using kinetics from previous literature and believed to be conservative and predicted very low rates of corrosion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Implementing Fuel Cladding Bonding and Assessing its Impact on Axial Gas Communication

Fuel rods irradiated in light-water reactors to burnup values above 45 MWd/kgU are subject to the formation of a chemical and mechanical bond between the fuel and cladding upon gap closure. The formation of the bond subsequently inhibits the ability of fission gases released from the fuel to flow freely to the plenum of the rod. The flowing of gases within fuel rods is referred to as axial gas communication. During transients, such as loss of coolant accidents, the bond may influence cladding deformation prior to breaking. Upon bond breakage, gases are able to more freely communicate to the lower pressure regions of the rod. The impact of bonding on gas communication and the ballooning behavior of the cladding during a loss of coolant accident is of interest to the nuclear industry in support of burnup extension for the existing light-water reactor fleet. In this report, a model to capture the effects of fuel cladding bonding in the BISON fuel performance code is presented. The theory of the model along with implementation testing is provided. A summary of a previously developed axial gas communication is given to set the stage for how the two models may be coupled together. A full-length pressurized-water reactor fuel rod demonstration is highlighted to evaluate the impact of including bonding on axial gas communication calculations using a preliminary coupling methodology. An overview of the next steps regarding the modeling of bonding, axial gas communication, and a more tightly coupled framework is also provided.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Characterization of Carbon Foam Products Made from Coal at Atmospheric Pressure

Carbon foam products are currently manufactured on a commercial scale utilizing a high-pressure autoclave in batch mode. In an effort to increase manufacturing speed and reduce cost, carbon foam products (both panels and aggregates) were produced at atmospheric pressure in a continuous belt kiln. Here, the microstructure, properties, and performance of these materials were characterized and assessed in various applications. Lightweight aggregate was explored for use as a proppant, as well as it use in concrete and asphalt. The thermal conductivity, water permeability, and fire resistance of panels were evaluated, and its performance with respect to screw pullout was compared against conventional building materials. Some interesting performance aspects and potential for future work was revealed. This work was supported by a cooperative agreement with the Department of Energy – National Energy Technology Laboratory (DE-FOA-0002185, Area of Interest 4: Coal-Derived Carbon Foam Produced via a Continuous Process, award number DE-FE-0031992).

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2025

Updates to the HTR-PROTEUS HALEU Benchmark Using Modern Analysis Methodologies

The HTR-PROTEUS IRPhEP Handbook benchmarks represent some of the highest quality benchmarks available for systems with TRISO-HALEU fuel, graphite pebbles, graphite reflector, and high neutron leakage. Nevertheless, significant variability in computed eigenvalue results was encountered in HTR-PROTEUS configurations that are relevant for transportation configurations. The team will investigate the source of these variabilities in the original benchmark. In addition, the team will investigate the HTR-PROTEUS subcritical measurements, kinetics data, and potential criticality effects from the introduction of hydrogen in the system (water ingress). The project will generate a benchmark evaluation of additional HTR-PROTEUS measurements that can significantly increase the value of these criticality benchmarks in testing nuclear codes and data to support transportation validation needs for industry and the U.S. NRC.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Radial–axial transport coordination enhances sugar translocation in the phloem vasculature of plants

Abstract Understanding mass transport of photosynthates in the phloem of plants is necessary for predicting plant carbon allocation, productivity, and responses to water and thermal stress. Several hypotheses about optimization of phloem structure and function and limitations of phloem transport under drought have been proposed and tested with models and anatomical data. However, the true impact of radial water exchange of phloem conduits with their surroundings on mass transport of photosynthates has not been addressed. Here, the physics of the Munch mechanism of sugar transport is re-evaluated to include local variations in viscosity resulting from the radial water exchange in two dimensions (axial and radial) using transient flow simulations. Model results show an increase in radial water exchange due to a decrease in sap viscosity leading to increased sugar front speed and axial mass transport across a wide range of phloem conduit lengths. This increase is around 40% for active loaders (e.g. crops) and around 20% for passive loaders (e.g. trees). Thus, sugar transport operates more efficiently than predicted by previous models that ignore these two effects. A faster front speed leads to higher phloem resiliency under drought because more sugar can be transported with a smaller pressure gradient.

58 GEOSCIENCES↗
DOE OSTI2026

Understanding the high-temperature behavior and corrosion resistance of Cr-Nb coated cladding for BWRs

While Cr-coated Accident Tolerant Fuel has proven successful in Pressurized Water Reactors, its dissolution in high dissolved oxygen Boiling Water Reactor environments represents a major technological barrier. Here, this study presents a comprehensive evaluation of novel Cr-Nb alloys as a coating breakthrough solution for Accident Tolerant Fuel cladding in Boiling Water Reactors. We systematically investigated Physical Vapor Deposited Cr-Nb coatings with 13% and 24 at% Nb content on commercial Zr-based cladding through rigorous testing under both normal and accident conditions. Our results demonstrate Cr-Nb coatings remained completely intact during extended Boiling Water Reactor autoclave testing, while pure Cr coatings failed catastrophically due to oxide dissolution. Under Loss-of-Coolant Accident conditions at 1100°C, Cr-Nb coatings maintained protective capability for 45 min. At 800°C, the alloy coatings matched pure Cr's excellent corrosion resistance. Despite microstructural changes during prolonged high-temperature exposure, no coating delamination occurred, ensuring continued protection of the underlying Zr substrate. These findings establish Cr-Nb coatings as a viable Accident Tolerant Fuel solution for Boiling Water Reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Flow Forecasting on New England’s Great River Hydro (CRADA 549) (Final Report)

Great River Hydro (GRH) is New England's largest conventional hydropower generator and operates sustainable hydropower systems in Vermont, New Hampshire, and Massachusetts. Several of GRH’s hydropower projects on the Connecticut River are largely constrained to operate as run-of-river; however, though a new FERC license they have opportunity in the form of flex hours where they are allowed to deviate operations to capture favorable market conditions or emergency operations. To maximize value from these new operations, GRH is considering new sources of forecasted inflow. Pacific Northwest National Laboratory (PNNL) is evaluating the quality of two commercial inflow forecasts for GRH under DOE Water Power Technology Office’s Notice of Opportunity for Technical Assistance: Improving Hydropower’s Value through Informed Decision Making. This project will provide an unbiased evaluation of the accuracy of inflow forecasts available through commercial vendors. GRH will use this information to support their decision on which inflow forecast to use under their new FERC license and to assess the overall value of the available inflow forecasting tools and practices. In this talk we will discuss details of the near real-time forecast evaluation process, the metrics involved in the evaluation, and some preliminary results. In addition, we’ll touch on the second phase of the project which involves evaluating scheduling strategies that identify windows of opportunity to change operations in ways that increase profit through participation in day ahead and real time markets.

13 HYDRO ENERGY↗
DOE OSTI2024

Synchronization and Rebound Effects in Residential Loads

Increasing fuel prices and capacity investment deferral place an increasing demand for peak reduction from distribution level systems. Residential and commercial devices, such as HVAC systems and water heaters, are increasingly involved in load control programs, and their use may generate synchronization and rebound effects, such as artificial peaks caused by device optimization. While there have been concerns over device synchronization, few studies quantify the extent of this effect with numerical values. In this study, we attempt to investigate whether control efforts result in device synchronization or rebound effects. We focus on three clustering methods – Ward’s clustering, Euclidean K-means, and Density-based spatial clustering of applications with noise – to evaluate the extent of synchronization of a fleet of water heaters and HVAC systems in Atlanta, Georgia. Our findings show that synchronization and rebound effects are present in the neighborhood’s water heaters, but none were found in the HVAC systems. Further, high usage water heaters are more susceptible to synchronization and rebound effects.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2021

Deactivation trends of Pd/SSZ-13 under the simultaneous presence of NO, CO, hydrocarbons and water for passive NO x adsorption

Pd-functionalized chabazite (Pd/SSZ-13) was evaluated for passive NO x adsorption (PNA) using low temperature combustion with diesel (LTC-D) reaction feed of the “United States Driving Research and Innovation for Vehicle efficiency and Energy sustainability” (U.S.DRIVE) protocol. Notably as per the protocol conditions, 12% O 2 + 6% CO 2 + 6% H 2 O was flown in all cases. NO-uptake studies in the presence of the LTC-D feed showed a systematic decline with (NO:Pd)molar changing from 0.5 to 0.4 after 10 trials. Control experiments showed more pronounced decline for NO + CO case, with appreciable intial (NO:Pd)molar value of 0.4 yet comparable-to-the-LTC-D-feed decline after 8 trials. CO-induced particle formation and larger extent of particle sintering was also evident from TEM analysis. Other controls did not exhibit trial-dependent deactivation as(NO:Pd)molar values were constant, at 0.3, 0.2, 0.5 and 0.2 for the NO, NO+H 2 , NO + unsaturated hydrocarbon (C 2 H 4 , C 3 H 6 ) and NO + saturated HC (C 3 H 8 , C 10 H 22 ) feeds respectively. In addition to these quantitative differences, desorption behaviors are qualitatively different. While only one major desorption event is observed for the full LTC-D, NO+ CO and NO+ unsaturated HC-controls, desorption occurs in two distinct stages for NO, NO + saturated HCs and initial-NO+H 2 trials. This arises due to inherent differences in Pd sites while exposed to these chemically distinct feeds. Furthermore, presence of reducing agents such as CO and unsaturated HCs in the feed result in almost complete elimination of lower, sub-200 °C desorption peak. The higher temperature desorption peaks, at > 300 °C is associated with NO strongly bound to ionic Pd sites and are prevalent under reducing conditions. Using DRIFTS also, three complexes leading to PNA are assigned, [O = N–Pd 2 +(OH)–Z], [O = N–Pd 2 +(Z2)] and [O = N–Pd 2 +(H 2 O)y–Z] with the latter being clearly observed upon water exposure. Pd/SSZ-13 showed higher hydrocarbon trapping than the SSZ-13 counterpart.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

UNF-ST&DARDS Enhancements for RCCA Data in As-loaded Dual Purpose Cask Models

This report summarizes the work performed to enable detailed modeling of rod cluster control assembly (RCCAs) in dual purpose canisters (DPCs) in as-loaded configurations using the Used Nuclear Fuel – Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). The goal of this project was to evaluate the reactivity impact that RCCAs have on k eff of DPCs with pressurized water reactor (PWR) fuel to potentially use the additional margin in future post-closure criticality safety analysis. This preliminary evaluation determined the number of DPCs that currently require compensatory actions prior to emplacement in a repository because of their high reactivity under post-closure criticality scenarios, which could be made acceptable by including the as-loaded RCCAs as specified in the Unified Database (UDB). This report briefly describes the modeling methods currently used within UNF-ST&DARDS and the modifications made to automate inclusion of the as-loaded RCCAs in the DPCs for post-closure criticality calculations for the loss of neutron absorber (NA) scenarios, or NA models. For the Zion site, the loss of basket scenarios, or degraded basket (DB) models, are also included. Discussion is also provided regarding the UDB data compared to available site-specific loading map data for the sites, which are specifically evaluated herein. This report compares the modifications to a similar evaluation for the Zion by site conducted by Walker as an initial validation. After the partial validation, all existing PWR sites with applicable NA models within the UDB were evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Seasonal forecasting skill for the High Mountain Asia region in the Goddard Earth Observing System

Seasonal variability of the global hydrologic cycle directly impacts human activities, including hazard assessment and mitigation, agricultural decisions, and water resources management. This is particularly true across the High Mountain Asia (HMA) region, where availability of water resources can change depending on local seasonality of the hydrologic cycle. Forecasting the atmospheric states and surface conditions, including hydrometeorologically relevant variables, at subseasonal-to-seasonal (S2S) lead times of weeks to months is an area of active research and development. NASA's Goddard Earth Observing System (GEOS) S2S prediction system has been developed with this research goal in mind. Here, we benchmark the forecast skill of GEOS-S2S (version 2) hydrometeorological forecasts at 1–3-month lead times in the HMA region, including a portion of the Indian subcontinent, during the retrospective forecast period, 1981–2016. To assess forecast skill, we evaluate 2 m air temperature, total precipitation, fractional snow cover, snow water equivalent, surface soil moisture, and terrestrial water storage forecasts against the Modern-Era Retrospective analysis for Research and Applications, Version 2 (MERRA-2) and independent reanalysis data, satellite observations, and data fusion products. Anomaly correlation is highest when the forecasts are evaluated against MERRA-2 and particularly in variables with long memory in the climate system, likely due to the similar initial conditions and model architecture used in GEOS-S2S and MERRA-2. When compared to MERRA-2, results for the 1-month forecast skill range from an anomaly correlation of R anom =0.18 for precipitation to R anom =0.62 for soil moisture. Anomaly correlations are consistently lower when forecasts are evaluated against independent observations; results for the 1-month forecast skill range from R anom =0.13 for snow water equivalent to R anom =0.24 for fractional snow cover. We find that, generally, hydrometeorological forecast skill is dependent on the forecast lead time, the memory of the variable within the physical system, and the validation dataset used. Overall, these results benchmark the GEOS-S2S system's ability to forecast HMA hydrometeorology.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Development and demonstration of a methodology to evaluate high burnup fuel susceptibility to pulverization under a loss of coolant transient

For economic reasons, the US nuclear industry is renewing efforts to build a technical basis to extend rod average burnup limits above the current regulatory burnup limit of 62 GWd/MTU. The primary driver is to increase pressurized water reactor cycle lengths to 24 months, reducing the number of fresh fuel assemblies and core design constraints, thereby making core energy utilization more efficient. However, fuel pellet fragmentation and pulverization, termed high burnup fuel fragmentation (HBFF), has been observed in the high burnup (>90 GWd/MTU) Halden loss-of-coolant-accident (LOCA) integral test series. The issue gained attention when fuel fragmentation and pulverization were also observed closer to the current US regulatory limit during the US Nuclear Regulatory Commission (NRC) sponsored out-of-core integral test at Studsvik Nuclear in early 2011. This led to NRC concerns with potential changes to fuel and core designs relative to fuel pellet pulverization. In a letter to the NRC Commissioners, the staff specifically identified a need to “…define the boundary of safe operation for key fuel design and operating parameters,” stating that “the staff is challenged to evaluate the acceptability of future fuel design advancements and fuel utilization changes.” As such, it can be concluded that HBFF and potential dispersal into the reactor coolant system introduces additional complications in light-water reactor (LWR) fuel safety evaluations. However, it is not clear how much fuel will be susceptible to HBFF; nor has there been a methodology developed to evaluate fuel susceptibility to HBFF. To that end, this paper proposes an analysis methodology to assess fuel susceptibility to HBFF during LOCA scenarios. The work presented here uses the BISON fuel performance code to evaluate a representative pressurized water reactor fuel rod exposed to a rod average burnup of 75 GWd/MTU. Sensitivity studies investigated the impact of the peak cladding temperature, transient fission gas released, and pre-transient fission gas release on cladding ballooning and burst timing. Subsequently, a methodology to assess fuel susceptibility to HBFF will be developed based on experimental data published in the open literature. The methodology will then be demonstrated by calculating the mass of fuel susceptibility to HBFF. The BISON results conclude that increasing peak cladding temperature drastically decreased time to failure, and decreased balloon size both of which have been confirmed experimentally. Additionally, the effect of pre-transient and transient fission gas release affected cladding balloon size and burst timing. Finally, fuel susceptibility to HBFF significantly decreased as a function of peak cladding temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Performance of sweeping gas membrane distillation for treating produced water: Modeling and experiments

Membrane distillation (MD) has shown promise for desalination of produced water (PW), but there are very few studies reported on the performance of MD with real PW and the characteristics of the fouling and scaling in MD. Among different MD configurations, sweeping gas MD (SGMD) is the least studied configuration. Here we report the application of SGMD for treating high-salinity PW (123 g L -1 ). The SGMD operating conditions were first carefully optimized through pure water MD experiments by varying operating conditions. The flux evaluation was accompanied by a detailed Aspen Custom Modeler simulation based on the electrolyte-NRTL thermodynamic model to better understand the heat and mass transfer in the feed and permeate channel. Supported by experiments, modeling results confirmed the partial condensation of the vapor in the permeate channel. Furthermore, it was concluded that the saturation of the sweeping gas is controlling the overall mass transfer through the membrane. Selected membranes and optimized conditions were then used to conduct short-term and long-term performance tests with PW. In long-term MD test, the concentration of the PW increased by a factor of two (50% water recovery), while the membrane exhibited small flux decline. Elemental analysis showed that strontium and sodium were the major scalant compounds.

42 ENGINEERING↗
DOE OSTI2021

The Role of Oxidation Resistance in High Temperature Alloy Selection for a Future with Green Hydrogen

Hydrogen is being considered integral to the future energy landscape but there is limited mechanistic understanding and a lack of predictive models to describe the combined effects of alloy and gas composition, temperature, thermal cycling and water vapor contents on the oxidation behavior of high-temperature materials. Experimental evaluations were combined with coupled thermodynamic-kinetic modeling to investigate the oxidation behavior of five representative Ni-based superalloys in two water vapor contents (10%, 60% H 2 O) under thermal cycling (1-h, 100-h cyles) conditions at 800°C and 1000°C. The alloy with the highest Ti content demonstrated the poorest cyclic oxidation behavior while the alloy with highest Cr and Al contents was expected to continue to support protective formation of a compact Al 2 O 3 scale. Finally, accelerated degradation of the chromia-forming alloys was observed in the higher water vapor content but the impact on transient oxidation of the alumina-forming alloys needs further investigations.

36 MATERIALS SCIENCE↗