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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Nondestructive Examination Guidance for Dry Storage Casks

In this report, an assessment of NDE methods is performed for components of NUHOMS 80 and 102 dry storage system components in an effort to assist NRC staff with review of license renewal applications. The report considers concrete components associated with the horizontal storage modules (HSMs) as well as metal components in the HSMs. In addition, the report considers the dry shielded canister (DSC). Scope is limited to NDE methods that are considered most likely to be proposed by licensees. The document, ACI 349.3R, Evaluation of Existing Nuclear Safety-Related Concrete Structures, is used as the basis for the majority of the NDE methods summarized for inspecting HSM concrete components. Two other documents, ACI 228.2R, Nondestructive Test Methods for Evaluation of Concrete in Structures, and ORNL/TM-2007/191, Inspection of Nuclear Power Plant Structure--Overview of Methods and Related Application, supplement the list with additional technologies that are considered applicable. For the canister, the ASME B&PV Code is used as the basis for NDE methods considered, along with currently funded efforts through industry (Electric Power Research Institute [EPRI]) and the U.S. Department of Energy (DOE) to develop inspection technologies for canisters. The report provides a description of HSM and DSC components with a focus on those aspects of design considered relevant to inspection. This is followed by a brief description of other concrete structural components such as bridge decks, dams, and reactor containment structures in an effort to facilitate comparison between these structures and HSM concrete components and infer which NDE methods may work best for certain HSM concrete components based on experience with these other structures. Brief overviews of the NDE methods are provided with a focus on issues and influencing factors that may impact implementation or performance. An analysis is performed to determine which NDE methods are most applicable to specific components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Performance Modeling of Materials-Based Hydrogen Storage Systems for Automotive Applications

One of the challenges of fuel cell vehicles is storage of the hydrogen fuel. In an effort to reduce the temperature and pressure limitations of storing neat hydrogen either as a high pressure gas or cryogenic liquid, material-based hydrogen storage is being developed. Hydrogen stored either as an adsorbent or as a chemical compound not only provides more reasonable pressures and temperatures but also improves the storage density of the hydrogen. Modeling tools have been developed as part of the U.S. Department of Energy's Hydrogen Storage Engineering Center of Excellence to assess the performance of these materials in light duty vehicle applications. The purpose of these modeling tools is to evaluate newly developed hydrogen storage materials and ensure that they meet customer expectations for driving range and performance before further investment. Recent advances in these modeling tools have been made to allow direct material property inputs. Such material properties as hydrogen capacity, isotherms and other thermodynamic and kinetic properties can reasonably be measured by materials researchers. Using these properties as inputs, the models can then provide estimates of the mass and volume of the storage system and the driving range and fuel efficiency of the vehicle during transient operation with four different drive cycles. The system sizing estimator and the framework model used to perform this analysis will be described and results presented for adsorbents and chemical based storage materials.

25 ENERGY STORAGE↗
DOE OSTI2017

Outcomes of the JNT 1955 Phase I Viability Study of Gamma Emission Tomography for Spent Fuel Verification

The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly has been assessed within the IAEA Support Program project JNT 1955, phase I, which was completed and reported to the IAEA in October 2016. Two safeguards verification objectives were identified in the project; (1) independent determination of the number of active pins that are present in a measured assembly, in the absence of a priori information about the assembly; and (2) quantitative assessment of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup, under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives was evaluated across a range of fuel types, burnups and cooling times, while targeting a total interrogation time of less than 60 minutes. The evaluations were founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. The simulated instrument response data were then processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics were defined and used to evaluate the performance of the methods.This paper describes the analysis framework and metrics used to predict tomographer performance. It also presents the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA. Finally, it gives examples of the expected partial-defect detection capabilities for some fuels and diversion scenarios, and it provides a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.

Jacobsson-Svard, Staffan↗
DOE OSTI2024

Towards a NEAMS-based high-fidelity model of the MARVEL reactor

This report outlines the progress of Idaho National Laboratory in developing a high-fidelity and high-resolution model of the Microreactor Applications Research Validation and Evaluation reactor. The model was developed under the Nuclear Energy Advanced Modeling and Simulation microreactor application driver at Idaho National Laboratory. The overarching objective of this activity is the development of a high-fidelity multiphysics MARVEL model using NEAMS tools, and to verify and validate NEAMS tools against MARVEL reference simulation and experimental data, respectively. This is a unique opportunity to conduct multiphysics analysis on a soon-to-be-deployed microreactor. This multiphysics model developed under the NEAMS-funded INL microreactor application driver leverages three single-physics models coupled via the MOOSE’s MultiApp and Transfer systems. The latter systems enable in-memory data transfer between MOOSE-based and MOOSE-wrapped applications. The first single-physics model, that functions as main application, leverages Griffin to model the neutron transport in the core through the discontinuous finite element (DFEM) discrete ordinates solver (SN). Several optimization flags that were developed by the Griffin developer team were beta-tested to enhance the solver’s performance. These include the combined use of using_average_xs and update_averaged_xs_on that enable to avoid expensive on-the-fly cross sections evaluations at each linear iterations in favor of evaluations of the macroscopic cross sections at each Picard iteration. The second single-physics model uses BISON to handle solid heat transfer and asymptotic hydrogen redistribution analysis in the fuel. While the model returns consistent results for the temperature and hydrogen distribution in the fuel, a mismatch was noticed in the calculated temperature in the reflector due to the value of the gap conductance used in our model. Ongoing investigations are being performed to assess the origin of this discrepancy. Finally, the System Analysis Module (SAM) was used to model the flow of the sodium-potassium eutectic in the primary loop. A first verification was also performed showing good agreement in terms of mass flow rate and inlet temperature. All mesh files were generated using the MOOSE Reactor module, removing the need for external meshing tools. Notably, this workscope represents one of the initial applications of the MOOSE Reactor module for modeling highly irregular geometries. The use of the reactor module significantly streamlined the mesh generation process. The full multiphysics mode, that combines all the single physics models, was leveraged to conduct initial steady-state multiphysics simulations to compute power, and temperature distribution in the reactor. Initial testing was performed for transient simulations as well. In this case, the new checkpoint restart capability for eigenvalue calculations was tested showing the capability for streamlined restart of transient calculations. Future work will focus on improving the fidelity of the model by performing comprehensive code-to-code comparisons. For instance, the full-core Griffin neutronics model will be benchmarked against MCNP reference results, that were provided by the MARVEL design team. Additionally, the SAM T/H model will be verified against reference RELAP-5 results for selected accident scenarios. Besides code-to-code verification exercises, the model fidelity will be improved by replacing the single-channel SAM model with a more complex SAM-Pronghorn coupled model, in which the sub-channel capability is deployed to obtain radial temperature resolution in the coolant. This model will be developed in synergy with the NEAMS thermal hydraulics team.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

NRC Multiphysics Analysis Capability Deployment (FY2021 - Part 2)

This report details the progress and activities of Idaho National Laboratory (INL) on the Nuclear Regulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.” Task 4b was completed for this report. INL developed a sample problem showing how to use Serpent 2 to calculate macroscopic cross sections for use in Griffin for a typical sodium fast reactor Unprotected Loss of Flow (ULOF) transient. The complicating factor is the relatively large axial gradient of the coolant (and hence fuel) temperature and its evolution during the transient. A 3D Griffin model of the Advanced Burner Test Reactor (ABTR) is coupled to the System Analysis Module (SAM) to perform the ULOF transient. The model includes various explicit feedback mechanisms for fast reactors, including Doppler, radial expansion from the displacement of the support plate, and axial expansion from the displacement of the fuel pins. The results for the ULOF transient are consistent with published values. We observed small effects in the transient results that arise from the superhomogenization equivalence correction of uniform and nonuniform temperature data sets, but they diminish as more dominant reactivity mechanisms are added to the model. Potential improvements to the neutronics model include: adding the differential control rod and sodium feedback mechanisms, implementing cross section re-homogenization based on the partial volume of the various materials, and enhancing the kinetics parameters. Furthermore, the fluids model only includes four subchannels to represent the entire active core due to performance limitations in SAM when attempting to model one subchannel per assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

RISMC Industry Application no. 1 (ECCS/LOCA) Core characterization automation: Reference PWR designs for IA no. 1

The U.S. NRC is currently proposing rulemaking '10 CFR 50.46c' to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that both cladding and ECCS performance need to be considered in a coupled manner. This implies that LOCA methodologies will need to be updated for regulatory purposes. The Risk-Informed Safety Margin Characterization (RISMC) Pathway, as part of the DOE Light Water Reactor Sustainability (LWRS) research and development program initialed several programs at the Idaho National Laboratory (INL) to support the industry by offering potential solutions to this new LOCA/ECCS analysis. One of these programs is the Industry Application no. 1 (IA1). It's goal is to develop an integrated evaluation model called LOCA Toolkit for the US (LOTUS) to revisit how risks are evaluated in a LOCA under 10 CFR 50.46c. The focus of LOTUS is to establish the automation interfaces among the physics disciplines involved, namely the core design automation, the fuels/clad performance, the system analysis, the risk assessment and the core design optimization. The risk-informed margins management approach for IA1 will provide a means of quantifying the impact on the key LOCA analysis figures-of-merit like peak cladding temperature (PCT), equivalent cladding reacted (ECR), etc. of a change in LOCA analysis inputs. The information obtained can then be used for decision-making and margin management. The modular approach for LOTUS will enable plant owners/vendors to further customize the LOTUS framework for use within their established codes and methods. One of the LOTUS modules deals with the core design automation (CD-A). The purpose of this module is to supply the LOCA analysis with initial conditions, i.e. assembly/pin power histories, power shapes, etc. to be employed by the Fuels/Clad Performance (EP) and System Analysis (SA) modules. In the CD-A module, the plant owner/operator will characterize his core design with the tools he has. For the current LOTUS demonstration, the tools used are the already coupled codes PHISICS and RELAP5-3D2,3. Cross section generation will be done using the Studsvik lattice code HELIOS-2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1991

Simulating The Performance Of A Scramjet

SCRAM computer program determines, according to one-dimensional mathematical model, cycle performance for airframe-integrated subsonic or supersonic combustion Ramjets. Reliable, efficient, and speedy ramjet designer's software tool usable on all standard computers down to those compatible with IBM PC-AT's. Scramjet cycle analysis code simulates, from nose to tail, hydrogen-fueled, airframe-integrated scramjet in real gas flow with equilibrium thermodynamic properties. Enables rapid scramjet performance estimates. Written in FORTRAN 77.

Walton, James T.↗
DOE OSTI2022

The international research reactor conversion effort and its contribution to the validation of reactor physics and thermal-hydraulics codes

Full text of publication follows. With about 200 reactors in operation, civilian Research and Test Reactors (RTRs) represent approximately a third of the global nuclear fleet. RTRs make use of core designs that are drastically different from commercial power plants to perform a wide variety of non-power applications that greatly benefit society. Because they often rely on high neutron fluxes, RTRs are designed with relatively compact cores and as a result, prior to the 1980's, were often deployed using Highly Enriched Uranium fuel (HEU, {sup 235}U/U = 20 wt. %). Due to proliferation concerns, the international community aims at eliminating the use of HEU in civilian facilities and favor instead the use of Low Enriched Uranium fuel (LEU, {sup 235}U/U < 20 wt. %). A program to support conversion of the world's RTRs to LEU fuel has been initiated in 1978 by the U.S. Department of Energy (DOE). This program is still alive today and has achieved more than 103 conversion metrics. Today, the program focuses heavily on the conversion of so-called high-performance RTRs, which are far more challenging than previous conversions as they require new fuel element designs and the use of new, higher density LEU fuel forms. Development and qualification of new LEU fuel elements is ongoing and requires extensive engineering analysis and testing. Both activities require the development and validation of codes and methods for reactor physics, thermal-hydraulics, and multi-physics, which in turn rely on experiments performed in RTRs or other experimental facilities. This talk will present the collection of RTR experimental data and benchmark analyses from the international conversion program that contribute to the validation of computer codes and methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Irradiation Thermo-Mechanical Modeling and Analysis of University of Missouri Research Reactor HEU Fuel Plates

The University of Missouri Research Reactor (MURR) located in Columbia, Missouri is one of six U.S. High Performance Research Reactors (USHPRR), including one critical facility, that is actively collaborating with U. S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply to convert from the use of highly enriched uranium (HEU; $\geqslant$ 20 wt% U-235) to low-enriched uranium (LEU; <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on an alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow the conversion to LEU of MURR, as well as four other USHPRR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Characterization of modeling and experimental data inconsistencies from burst testing for high-burnup commercial fuel rod applications

Fuel fragmentation, relocation, and dispersal are some of the largest issues remaining in the nuclear industry before rod-average burnup can be increased beyond 62 GWd/tU. The issue is primarily related to the potential for fuel to be dispersed into the reactor primary system, which may increase public risk. One way to prevent dispersal is to avoid cladding burst. The objective of this work is to support the high burnup safety case by evaluating cladding burst under high-burnup, full-length fuel rods and to identify uncertainties that could improve model predictions. The results of this analysis will evaluate realistic, prototypic loss-of-coolant accident (LOCA) conditions; support future cladding burst test designs; and inform the development of mechanistic material models. Realistic high-burnup operating conditions were implemented in the BISON fuel performance code to simulate steady-state and LOCA transient fuel rod evolution to the point at which cladding burst occurred. Parametric studies are performed to assess code response to changes in rod internal pressures and heating rates. Results were compared with simulated LOCA experiments to identify inconsistencies between commercial fuel rod analysis and experimental validation. The representative full-length fuel rod LOCA simulation results did not agree with cladding burst tests. Cladding burst tests indicated burst occurring well below (100–150 °C) those calculated in the full-length fuel rod LOCA analysis. Further investigation indicated that the cladding burst tests do not appear to be representative for full-length fuel rods. The inconsistency investigated in this work showed that the differences between the BISON simulation and the experiment's cladding burst conditions arise from an incomplete characterization of the cladding surface temperature, detailed rodlet characterization, lack of cladding strain measurements, and uncertainty in the cladding creep and failure models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Robust measurement of microbial reduction of graphene oxide nanoparticles using image analysis

ABSTRACT Shewanella oneidensis ( S. oneidensis ) has the capacity to reduce electron acceptors within a medium and is thus used frequently in microbial fuel generation, pollutant breakdown, and nanoparticle fabrication. Microbial fuel setups, however, often require costly or labor-intensive components, thus making optimization of their performance onerous. For rapid optimization of setup conditions, a model reduction assay can be employed to allow simultaneous, large-scale experiments at lower cost and effort. Since S. oneidensis uses different extracellular electron transfer pathways depending on the electron acceptor, it is essential to use a reduction assay that mirrors the pathways employed in the microbial fuel system. For microbial fuel setups that use nanoparticles to stimulate electron transfer, reduction of graphene oxide provides a more accurate model than other commonly used assays as it is a bulk material that forms flocculates in solutions with a large ionic component. However, graphene oxide flocculates can interfere with traditional absorbance-based measurement techniques. This study introduces a novel image analysis method for quantifying graphene oxide reduction, showing improved performance and statistical accuracy over traditional methods. A comparative analysis shows that the image analysis method produces smaller errors between replicates and reveals more statistically significant differences between samples than traditional plate reader measurements under conditions causing graphene oxide flocculation. Image analysis can also detect reduction activity at earlier time points due to its use of larger solution volumes, enhancing color detection. These improvements in accuracy make image analysis a promising method for optimizing microbial fuel cells that use nanoparticles or bulk substrates. IMPORTANCE Shewanella oneidensis ( S. oneidensis ) is widely used in reduction processes such as microbial fuel generation due to its capacity to reduce electron acceptors. Often, these setups are labor-intensive to operate and require days to produce results, so use of a model assay would reduce the time and expenses needed for optimization. Our research developed a novel digital analysis method for analysis of graphene oxide flocculates that may be utilized as a model assay for reduction platforms featuring nanoparticles. Use of this model reduction assay will enable rapid optimization and drive improvements in the microbial fuel generation sector.

Bennett, Danielle T. (ORCID:0009000188748827)↗
DOE OSTI2016

Assessment of SFR Wire Wrap Simulation Uncertainties

Predictive modeling and simulation of nuclear reactor performance and fuel are challenging due to the large number of coupled physical phenomena that must be addressed. Models that will be used for design or operational decisions must be analyzed for uncertainty to ascertain impacts to safety or performance. Rigorous, structured uncertainty analyses are performed by characterizing the model’s input uncertainties and then propagating the uncertainties through the model to estimate output uncertainty. This project is part of the ongoing effort to assess modeling uncertainty in Nek5000 simulations of flow configurations relevant to the advanced reactor applications of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Three geometries are under investigation in these preliminary assessments: a 3-D pipe, a 3-D 7-pin bundle, and a single pin from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility. Initial efforts have focused on gaining an understanding of Nek5000 modeling options and integrating Nek5000 with Dakota. These tasks are being accomplished by demonstrating the use of Dakota to assess parametric uncertainties in a simple pipe flow problem. This problem is used to optimize performance of the uncertainty quantification strategy and to estimate computational requirements for assessments of complex geometries. A sensitivity analysis to three turbulent models was conducted for a turbulent flow in a single wire wrapped pin (THOR) geometry. Section 2 briefly describes the software tools used in this study and provides appropriate references. Section 3 presents the coupling interface between Dakota and a computational fluid dynamic (CFD) code (Nek5000 or STARCCM+), with details on the workflow, the scripts used for setting up the run, and the scripts used for post-processing the output files. In Section 4, the meshing methods used to generate the THORS and 7-pin bundle meshes are explained. Sections 5, 6 and 7 present numerical results for the 3-D pipe, the single pin THORS mesh, and the 7-pin bundle mesh, respectively.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Fuel Cycle Research and Development Accident Tolerant Fuels Series 1 (ATF-1) Irradiation Testing FY 2016 Status Report

This report contains a summary of irradiation testing of Fuel Cycle Research and Development (FCRD) Accident Tolerant Fuels Series 1 (ATF 1) experiments performed at Idaho National Laboratory (INL) in FY 2016. ATF 1 irradiation testing work performed in FY 2016 included design, analysis, and fabrication of ATF-1B drop in capsule ATF 1 series experiments and irradiation testing of ATF-1 capsules in the ATR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Cooperative Research and Development Agreement between National Energy Technology Laboratory and Electric Power Research Institute (EPRI) [Abstract]

Carbon dioxide (CO 2 ) capture from flue gas generated by fossil fuel-fired power plants has been proposed as an efficient approach to limit CO 2 emissions to the atmosphere. Due to the high cyclic CO 2 sorption capacity, well-tuned adsorption chemistry and non-volatility, solid sorbents are widely studied for CO 2 capture processes such as pressure swing adsorption (PSA) and temperature swing adsorption (TSA). Realistic applications of traditional solid sorbent systems face many challenges since moisture and heat management are problematic and solid-solid heat exchange is inefficient. Recently, a novel solid sorbent system, the sorbent polymer composite (SPC), has been developed to overcome those challenges in an energy-saving CO 2 capture process using TSA membrane contactors. A SPC material is comprised of a powdered solid sorbent embedded into a hydrophobic and porous polymer matrix. It allows gases to permeate through and achieve full contact with the sorbents while rejecting water moisture in the CO 2 capture process. In this collaborative research, NETL will conduct performance testing of a novel SPC material and EPRI will conduct a cost analysis based on that performance data.

20 FOSSIL-FUELED POWER PLANTS↗
NASA NTRS1982

Measurements of the Fuel Mileage of a KC-135 Aircraft with and Without Winglets

The KC-135A Winglet Flight Research and Demonstration Program was a joint effort of the Air Force, NASA and the Boeing Military Airplane Company to flight test winglets on the KC-135A. The primary objective of the program was to verify the cruise performance improvements predicted by analysis and wind tunnel tests. Flight test data were obtained for winglets positioned at 15 deg cant/-2 deg incidence, 0 deg cant/-4 deg incidence, 15 deg cant/-4 deg incidence and for winglets off (baseline). Both fuel mileage and drag measurements were obtained. The 15 deg cant/-4 deg incidence winglet configuration provided the greatest performance improvement. The flight test measured fuel mileage improvement for a 0.78 Mach number was 3.1 percent at 8 x 10(5) pounds W/delta and 5.5 percent at 1.05 x 10(6) pounds W/delta. Correcting the flight measured data for surface pressure differences between wind tunnel and flight resulted in a fuel mileage improvement of 4.4 percent at 8 x 10(5) pounds W/delta and 7.2 percent at 1.05 x 10(6) pounds W/delta. The agreement between the fuel mileage and drag data was excellent.

Temanson, G. E.↗
NASA NTRS1996

Advanced Models for Aeroelastic Analysis of Propulsion Systems

This report describes an integrated, multidisciplinary simulation capability for aeroelastic analysis and optimization of advanced propulsion systems. This research is intended to improve engine development, acquisition, and maintenance costs. One of the proposed simulations is aeroelasticity of blades, cowls, and struts in an ultra-high bypass fan. These ducted fans are expected to have significant performance, fuel, and noise improvements over existing engines. An interface program was written to use modal information from COBSTAN and NASTRAN blade models in aeroelastic analysis with a single rotation ducted fan aerodynamic code.

Keith, Theo G., Jr.↗
DOE OSTI2015

Preliminary Physics Analysis of the Accident Tolerant Fuels Loop Experiment

After the Fukushima Daiichi power plant disaster in 2011, the United States Congress ordered the Department of Energy (DOE) to investigate new fuel and cladding types that would be more tolerant in an accident similar to the accident at Fukushima. The resulting program was dubbed the Accident Tolerant Fuels (ATF) program. As a part of this program, insertion is planned for an experiment to go into an ATR loop, and physics analysis must be performed. Discussed are the test fission power, void worth, and hardware failure reactivity insertion. For the sensor qualification test, the test fission power was 0 kW, the void worth was 0.42$, and the hardware failure was 0.01$ and 0.04$. For the fueled experiment, the safety case fission power was 170 kW, the void worth was 0.34$, and the hardware failure was 0$ and 0.03$.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

PNNL's Intermediate Characterization Summary for the MP-1 Experiment

Characterization of the as-fabricated fuel is being performed at Pacific Northwest National Laboratory (PNNL) in accordance with Characterization Plan for the Fabrication of U-10Mo for the MP-1 Experiment, hereinafter the MP-1 Characterization Plan. Similar characterization work is also being performed at two other organizations, INL and Los Alamos National Laboratory, to provide a detailed understanding of the as-fabricated fuel that would be irradiated in the MP-1 experiment. Under the MP 1 Characterization Plan, fuel foils with different fabrication parameters (such as rolling condition, rolling thickness reduction, co-rolling with Zr layers) are being studied. Similar samples from master foils were sent to each of the three organizations, so that testing and analysis can be performed independently using similar equipment and a standardized set of measurement and analysis procedures. A final consolidated report will be prepared based on this work and it will summarize all of the information obtained from different labs. The MP-1 experiment will provide an opportunity to understand the effects of processing conditions on the final fuel microstructure, to compare results obtained independently, and achieve a three-way validation. Currently, PNNL characterized 28 U-10Mo samples, by sectioning six pieces/specimens from each sample/foil, in accordance with the MP-1 Characterization Plan. These 28 samples consist of six types of MP-1 characterization foils from BWXT: 0.047 in. thick hot-rolled samples with and without Zr layers; 0.02 in. thick cold-rolled and annealed samples with and without Zr layers, and 0.01 in. thick cold-rolled and annealed samples with and without Zr layers. During the first year, PNNL completed sectioning of two subsamples per sample/foil and performed the necessary characterization on all 28 U-10Mo samples. This report covers the results of the first year’s analyses of PNNL’s MP-1 fuel characterization samples. Microstructure, chemical composition, carbide morphology, U-10Mo foil thickness, Zr thickness, mechanical properties (microhardness), and density were evaluated in both longitudinal and transverse directions for foils of the three different thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗