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At least 1,225 records · Page 68
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Status PIE Report on Legacy EBR-II and FFTF Metallic Fuel Experiments

A long term research focus of the US Department of Energy (DOE) Advanced Fuels Campaign (AFC) is to investigate technologies that can allow for increased actinide utilization in nuclear fuel for fast neutron spectrum reactors. In order to reach these goals, many experimental campaigns have been tested in an approximated fast reactor neutron spectra at the Idaho National Laboratory Advanced Test Reactor (ATR) in order to study the performance of metallic, nitride, ceramic and advance fuel forms. In addition and complementary to ATR testing, systematic characterization of experiments irradiated in true fast reactors are performed in order to assess ATR in qualifying fast reactor fuel. For this purpose, complementary postirradiation examinations (PIE) are performed on several metallic fuel pins previously irradiated in the EBR II. Advanced PIE techniques developed since the initial PIE of EBR II samples in the 1980’s have been applied in new examinations of irradiated samples during past EBR-II operations. The EBR-II metallic fuel pins analyzed in this report are X501 G591, an early minor actinide-bearing metallic fuel alloy; X430-T653, a U-19Pu-10Zr metallic fuel alloy with a larger diameter compared to EBR-II driver fuel, and more prototypic of future commercial fast reactors; X521-G594, a unique experiment to test the feasibility to burn weapon grade Pu without reprocessing; X496 CL47&CL39 from a long-life metallic fuel experiment with the goal to reach very high burnup; and X486-J555 a U-10Zr Mk-IIIA qualification fuel test that will serve as first of a kind “modelling driven PIE.” These legacy pins were evaluated against the historical fuel performance of previously irradiated fuel from literature. Fuel performance was investigated through a suite of PIE techniques including neutron radiography, gamma spectrometry, dimensional inspection, fission gas release measurements, chemical burnup analysis, optical microscopy, and advance microscopy. So far, the PIE campaign revealed fuel performance and characteristics that are generally in line with literature or EBR-II experience. Some exceptions are present, and they are directly discussed in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

MCNP AND CFD MODELING OF A SODIUM FAST REACTOR SUB-ASSEMBLY CHANNEL TO CAPTURE LOCALIZED TEMPERATURE PEAKING

Typical evaluation tools for sodium-cooled fast reactors (SFR) use simplified neutronic/thermal models to efficiently compute the power and temperature distributions within the core. However, their modeling capability can be limited in extreme cases such as when sharp inter-assembly power gradients occur, or for non-standard SFR assembly geometries. Higher fidelity tools that rely on Monte Carlo methods for neutron transport, and Computational Fluid Dynamics (CFD) for thermal analysis, are better suited in these instances. The higher accuracy comes at the cost of higher computational expense. MCNP and STAR CCM+ were selected for this analysis. While MCNP is capable of generating pin-level power distributions for a whole-core SFR model, it is prohibitively expensive to then conduct CFD analysis for every single assembly. A semi-empirical model was therefore developed to rapidly estimate pin temperature distributions throughout the whole core. It relies on a geometric correction factor for turbulent flow-mixing. The correction factor is obtained from a CFD simulation of a single fuel assembly with a uniform power distribution and an adiabatic boundary condition. Without using CFD, each sub-channel outlet temperature is calculated separately using 1-D heat-balance and Nusselt number correlations and the MCNP calculated pin-powers. The rudimentary 1-D model is then corrected for the flow-mixing effect based on the CFD data. The temperature calculation is performed by a code written in the Python language called CFD Monte-Carlo Developed Reconstruction of Sub-channel at a Quasi-Representative Level (CMDR-SQRL). CMDR-SQRL was shown to adequately predict CFD evaluated temperature distribution for a wide range of SFR configurations, including an extreme case with fuel assemblies nearby containing moderating material. The agreement demonstrates how geometric and power-driven effects can be decoupled within an SFR assembly. CMDR-SQRL is intended to be mainly used for scoping studies to provide core-wide evaluations efficiently.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Mode I fracture testing fixture

An apparatus and method for preparing Double Cantilever Beam (DCB) specimens are disclosed as an apparatus and method for conducting Mode I fracture resistance testing using the DCB specimens. In a first embodiment, a drill jig is used to locate the DCB specimen and guide a drilling process during creation of at least one through-hole in the DCB specimen. The drilling process may employ a traditional drill and drill bit, a laser drill, or a water jet. In another embodiment, a set of rotating pin blocks, each with a full-round or a half-round specimen pin at one end and a hanger full-round pin at the other end, engage the DCB specimen and facilitate the internal application of a fracturing load to the DCB specimen for the Mode I fracture resistance test. The present invention may significantly reduce the time and materials needed to prepare and test a DCB specimen.

Briggs, Timothy↗
DOE OSTI2022

High-voltage push to mate electrical interconnect

A high-voltage self-aligning push to mate electrical interconnect device is provided. The electrical interconnect device includes a housing with a mating end and a receiving end. The interconnect also includes an insulator bushing configured to be affixed to the receiving end of the housing. The interconnect also includes a center pin disposed within the housing. The center pin extends from the mating end of the housing to the receiving end of the housing. The interconnect also includes a connector socket disposed within the insulator bushing and which is in electrical communication with the center pin. The interconnect also includes an alignment sleeve including a chamfered mating funnel defining a void configured to accept and surround the housing and at least a portion of insulator bushing.

Martinez, David K.↗
DOE OSTI2022

Shrouded screw device for removing and depositing material

A shrouded screw apparatus includes an outer tube, a telescoping inner tube, an auger bit, and an anti-rotation pin. The outer tube has a first sidewall defining a first interior chamber. The first sidewall has an opening defined therethrough. The inner tube is positioned in the first interior chamber. The inner tube has a second sidewall defining a second interior chamber. The second sidewall has a longitudinal groove defined therein. The inner tube is configured to extend from and retract within the first interior chamber of the outer tube. The auger bit is positioned in the second interior chamber and is rotatably coupled to the inner tube. The anti-rotation pin is coupled to the first sidewall of the outer tube and extends inwardly into the first interior chamber. The pin engages the longitudinal groove of the inner tube.

47 OTHER INSTRUMENTATION↗
DOE OSTI2013

Optimization of multi-group cross sections for fast reactor analysis

The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO{sub 2}-UO{sub 2} with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Mode I fracture testing fixture

An apparatus and method for preparing Double Cantilever Beam (DCB) specimens are disclosed as an apparatus and method for conducting Mode I fracture resistance testing using the DCB specimens. In a first embodiment, a drill jig is used to locate the DCB specimen and guide a drilling process during creation of at least one through-hole in the DCB specimen. The drilling process may employ a traditional drill and drill bit, a laser drill, or a water jet. In another embodiment, a set of rotating pin blocks, each with a full-round or a half-round specimen pin at one end and a hanger full-round pin at the other end, engage the DCB specimen and facilitate the internal application of a fracturing load to the DCB specimen for the Mode I fracture resistance test. The present invention may significantly reduce the time and materials needed to prepare and test a DCB specimen.

Briggs, Timothy↗
DOE OSTI2015

The use of gamma spectrometry in the determination of fission products migration in irradiated fuel

Non destructive examinations realized in hot cells of LECA STAR facility give main data on irradiated fuel rods and pins. Among those examinations, gamma spectrometry allows access to fuel, inside the cladding, thanks to gamma rays of fission products such as {sup 137}Cs, {sup 154}Eu coming from pellets... From those gamma scannings we can detect the position and the length of the fuel column and of its pellets, calculate local burnup.. In the database of our lab we have already such gamma scannings on hundreds of rods or pins with different fuels, claddings and irradiations conditions (under nominal or non nominal). We have detected that in specific cases, an unusual shape of the {sup 137}Cs scanning (distribution quite different from those of {sup 154}Eu) can be explained by the migration of this isotope, moving to the cold sides of the pellet. This phenomenon is mainly associated with an increasing of the pellet's temperature. Based on our observations, we have developed a quantitative approach of the changes on the {sup 137}Cs scannings through the calculation of appropriate 'indicators'. Those calculations allow us to be able to localize, quantify and compare the {sup 137}Cs migration all along rods or pins. Those migration results are so quickly and easily achievable from gamma measurements and can then be easily correlated to other observations realized with destructive examinations, puncturing and calculations, already realized or to come. Because such a migration is the result of temperature increasing in the pellets, our indicators can be directly associated with this local temperature. In perspective, thermically activated phenomena such as geometrical changes in the shape of the pellets, fission gas release... can also indirectly be deduced from our indicators. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Determination of critical experiment correlations for experiments involving arrays of low-enriched fuel rods

Quantifying correlations among critical experiments has become a topic of interest over the last 10-15 years. Correlations arise from the shared use of fissile material, non-fissile materials, fixtures, and procedures in multiple experiments or multiple experiment series. A Monte-Carlo sampling technique proposed as a viable technical method to evaluate these correlations has been implemented in the SCALE 6.2 Sampler sequence. This paper presents an investigation of the use of this new tool and technique for low-enriched uranium pin array experiments with a thermal neutron energy spectrum. The Nuclear Energy Agency's Expert Group on Uncertainty Analyses for Criticality Safety Assessments has been examining methods of calculating the correlation coefficients among experiments from the LCT-007 and LCT-039 series. Results for two scenarios are presented to demonstrate the impact of pin pitch uncertainty modeling on the correlation coefficients. The convergence of the correlation coefficients is also investigated. Additional studies examine the impact of the stochastic uncertainty and the use of continuous-energy Monte-Carlo for individual realizations on the correlation coefficients. Correlation coefficients are also determined for the LCT-042 experiment series. This experiment series differs from other experiments because unique poison panels and fuel array separation distances may reduce correlation of the individual experiments within the series. The effects of pin pitch uncertainty and stochastic uncertainty are also examined for this series. Comparing the results of the two sets of experiments shows potential conclusions that can be drawn regarding viability of the Monte-Carlo sampling technique for determining critical experiment correlations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Nek5000 Large-Eddy Simulations for Thermal-Hydraulics of Deformed Wire-Wrap Fuel Assemblies

A well-characterized validation program has been defined for the thermal-hydraulic CFD modeling of 61-pin hexagonal packed fuel assemblies with helically wire-wrapped pins for both non-deformed beginning-of-life geometries and deformed later-in-life geometries typical of sodium-cooled fast reactors. The motivation for this work is the need to validate CFD codes to accurately model the flow and thermal performance of wire-wrapped fuel assemblies, particularly the deformed later-in-life bundles. The effect of pin and duct deformation is increasingly important as reactor designs look toward longer assembly lifetimes. As a part of a larger collaborative project with TerraPower, Texas A and M University (TAMU), and AREVA, this validation work consists of CFD simulations of four con-current experiments. Pre-test numerical simulations corresponding to each of the experiments has been conducted and are planned to be followed by post-test simulations that will incorporate test information as it becomes available to account for any as-built differences in geometry of test section and its specifications including the differences in boundary conditions of the actual tests. This paper focuses on Nek5000 LES results for the isothermal/TAMU undeformed configuration as a first step toward and initial conditions for the deformed isothermal/TAMU and/or conjugate heat transfer/AREVA cases. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Towards a Neutronic and Thermal-Hydraulic Safety Analysis of PWR Fuel Bow with SIMULATE5 and COBRA-TF

In the 1990's, collective fuel assembly bow was observed in the Ringhals PWRs in Sweden. Back then fuel vendors reinforced the fuel assemblies' dashpot, decreased hold-down spring forces, increased guide tube wall thickness, and introduced guide tube materials with higher creep resistance in order to improve dimensional stability. Nevertheless, twenty years later in 2010, the same kind of collective bow with S and C shape patterns started to appear in most of Germany's PWRs. At this time the SIMULATE5 (S5) core simulator with its delta-gap bow model (DGA) which was originally developed for safety analysis at Ringhals, could be used as an off-the-shelf solution to assess the impact of bow on Fq and FH. However, these two quantities are of secondary importance in German PWRs in normal operation because the most restricting licensing criteria set limits on DNBRmin values. For this purpose, a method was developed to extract S5 pin powers from DGA calculations and automatically generate COBRA-TF input files (CTF) to analyze DNBR values in the cores. In the following two sections, it is shown how the S5 model for DGA calculations has been verified with independent MCNP6 calculations in order to check its predictions on pin power changes for delta gaps up to 1 cm in size and how the existing licensed DNBR calculation methodology has been updated to bowed fuel assembly situations with the help of CTF. Finally, the applicability of the Doroschuk CHF correlation, which is the licensing standard in most German PWRs, is discussed. For the fuel bow patterns considered, it is assumed that only the gap between fuel assemblies change and that the pitch between fuel rods inside the assemblies remains largely unchanged. This approximation can be justified by the visual inspections conducted at the plants during the outages. Every calculation of the pin-power impact of delta-gaps in the core under operating conditions is only as good as the knowledge of the bow pattern. Measurements of fuel assembly bow at cold conditions have been made for some German PWRs for all the fuel assemblies in the core during plant outages. Together with methods developed by the fuel assembly vendors, a best-estimate bow pattern during operation could be reconstructed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Advances in Metallic Fuel Database Development and Data Qualification

The Fuels Irradiation and Physics Database (FIPD [1]) is a comprehensive repository of data and documents related to Uranium-Zirconium based metallic fuel test pins. This database stores operational conditions of these pins, calculated using a suite of Argonne National Laboratory analysis codes developed during the Integral Fast Reactor (IFR) program. Key calculated data include axial distributions of power, temperature, fluence, burnup, and isotopic densities. Additionally, the FIPD holds post-irradiation examination (PIE) data such as fission gas release, gas chemistry measurements, and axial distributions derived from profilometry, gamma scanning, and neutron radiography. Complementing these data is an extensive archive of documents related to various pins and experiments. These include raw PIE records, design details, safety analyses, and operational reports. More detail about FIPD can be found in ref. [2]. The database development is an ongoing effort covering metallic fuel experiments from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The recent improvements to the database and the data QA status are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Data Acquisition System Selection and Calibration of Resistive Moisture Content Measurements for Large-Scale Field Studies in Cold Climate Residential Building Envelope Performance

The residential building stock built before the energy codes were enforced has several significant inefficiency problems in terms of insulation and air leakage. To decrease these inefficiencies, building retrofits are necessary. However, if the envelope is not appropriately designed, excessive accumulation of moisture content and thus mold formation and decay inside the envelope layers can be a vital problem. This risk becomes higher, especially in extreme climate conditions such as cold winters and hot and humid summers as in some northern regions of the U.S. Field studies are essential to test the long-term hygrothermal performance of building envelopes. Although in-situ temperature, RH, and heat flux measurements are straightforward, moisture content measurements are cumbersome. Mainly, because of the heterogeneous nature of the wood materials, deviations and nonuniformities within the materials are unavoidable. Resistance measurements are one of the oldest methods used to measure the moisture content of wood and other building materials. In large-scale studies, it is commonly preferred to use multi-purpose data acquisition systems (DAQ) and custom-made or prefabricated moisture pins to measure the electrical resistance (and thus moisture content) of critical building materials. These multi-purpose DAQ systems generally provide lower costs and offer more flexibility. However, these systems require calibration and fine-tuning to achieve accurate moisture content measurements. A large-scale, two-year-long field study was conducted in northern Minnesota to monitor the hygrothermal performance of residential retrofit wall systems in cold climates. Two base case walls and sixteen different wall treatments were tested. Moisture contents were measured at various layers in each wall treatment using 85 sets of moisture pins. This paper focuses on the overall approach, fabrication, and calibration methodology for the combination of custom-made moisture pins and a multi-purpose DAQ. The aim is to directly use the low-excitation multi-purpose DAQ without any extra voltage regulator. A half-bridge circuit is used to measure wood resistance with 4V excitation voltage and 100 kΩ and 500 kΩ reference resistors. The system is calibrated for four different materials: Douglas fir, lodgepole pine, western red cedar, and oriented strand board (OSB). Calibration experiments were done under controlled conditions in 50% and 65% RH test chambers. Resistance-based moisture content calibration curves are obtained for each species. Results show that higher reference resistors provided better calibration curves for lower excitation voltages.

Desjarlais, Andre Omer↗
DOE OSTI2024

Fuel Performance Evaluation of THOR-C Experiments

The Temperature Heatsink Overpower Response Commissioning (THOR-C) and THOR-Metal (THOR-M) experiments will be performed as part of an ongoing project for testing sodium fast reactor fuels with the Japan Atomic Energy Agency (JAEA). The THOR-C experiments consist of fresh metallic fuel pins and have been analyzed using the ABAQUS, Ansys codes and the BISON fuel performance code. THOR-M-Loss of Flow-1 (THOR-M-LOF-1) is designed to test an EBR-II irradiated fuel pin under LOF conditions. Simulation of the THOR-MLOF-1 experiment required first simulating the base irradiation of the fuel pin in EBR-II. MFUEL module of SAS4A/SASSYS-1 [1] is a physics-based metallic fuel performance model applicable to the normal operation, transient scenarios and fuel failure modeling including scenarios with bulk fuel melting. The model has been validated using EBR-II normal operation, separate effect transient tests as well as TREAT M-Series transient tests [2]. In this study, MFUEL models has been utilized together with a new capsule heat transfer model developed in this project. The new heat transfer model was necessary due to (1) significant amount of heat losses that required 2D heat transfer, (2) the presence of a titanium heat sink, rejecting a significant amount of heat, and (3) stagnant coolant conditions, which are inconsistent with SAS4A/SASSYS-1 (SAS) heat transfer model. Updates to SAS4A/SASSYS-1 and MFUEL has been described below, followed by a preliminary validation effort using the results from THOR-C-2 fresh fuel capsule experiment. A previous study for THOR-C-2 analysis using BISON code is also utilized in this study to model this test [3]. [1] D. O’Grady, A. J. Brunett, L. Ibarra, A. Karahan, T. Kim, T. S. Sumner, R. Thomas, T. H. Fanning, “The SAS4A/SASSYS-2 Version 5.7 Safety Analysis Code System,” Argonne National Laboratory,ANL/NSE-SAS/5.7, (2023). [2] A. Karahan, T. Kim, T. Fanning, D. O’Grady, “Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1,” Argonne National Laboratory, ANL/NSE-23/11, (2023). [3] M. Mihelish, A. Zabriskie, K. Paaren, P. Medvedev, C. Jensen, “Fuel Performance Predictions for the TREAT THOR-C Experiments,” Idaho National Laboratory, INL/RPT-23-73397, Revision 0, (2023)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Shrouded screw device for removing and depositing material

A shrouded screw apparatus includes an outer tube, a telescoping inner tube, an auger bit, and an anti-rotation pin. The outer tube has a first sidewall defining a first interior chamber. The first sidewall has an opening defined therethrough. The inner tube is positioned in the first interior chamber. The inner tube has a second sidewall defining a second interior chamber. The second sidewall has a longitudinal groove defined therein. The inner tube is configured to extend from and retract within the first interior chamber of the outer tube. The auger bit is positioned in the second interior chamber and is rotatably coupled to the inner tube. The anti-rotation pin is coupled to the first sidewall of the outer tube and extends inwardly into the first interior chamber. The pin engages the longitudinal groove of the inner tube.

Kontras, Evan A.↗
DOE OSTI2025

SCM overview & the EBR-II shutdown heat removal tests validation.

Pronghorn is an engineering-scale, coarse-mesh, thermal-hydraulics tool for supporting reactor-core simulations of advanced nuclear reactors. Most of the current efforts in Pronghorn have been devoted in developing porous finite-volume capabilities and adapting closure correlations for coarse-mesh thermal-hydraulics modeling. However, for liquid-metal reactors (LMRs) with wire-wrapped fuel pin assemblies, a pin-level thermal-hydraulic resolution is required for most safety case studies (pin rupture, channel blockage, etc.). For this purpose, a new Subchannel application is developed in MOOSE, which affords the required flow field resolution, while still preserving an engineering-scale approach. This new solver can be natively coupled to Pronghorn and other MOOSE objects to enable full-core, multi-physics, multi-scale engineering studies. This presentation presents the main features of the SCM code and demonstrates a validation case based on the EBR-II SHRT tests.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

CAPTAN+X Data Converter Integration

Fermi National Accelerator Laboratory's CAPTAN (Compact And Programmable daTa Acquisition Node) series provides a flexible hardware platform for data acquisition across a range of experiments and facilities. The latest iteration, CAPTAN+X, is built around a Kintex-7 FPGA supporting four FPGA Mezzanine Card (FMC) connections. As part of a broader laboratory effort to bring facility systems under a Model-Based Systems Engineering (MBSE) framework, CAPTAN+X is one of several systems slated to be incorporated into this modeling environment in the near term. A necessary step toward that goal is incorporating the platform's core functionality, which centers on integration with the LXD31K4 FMC, a data converter module combining dual AD9652 analog-to-digital converters and dual AD9142A digital-to-analog converters. Achieving compatibility required resolving pin-mapping conflicts between the LXD31K4's High Pin Count connector and the CAPTAN+X's available pin types, adapting a Board Support Project originally written for an UltraScale-class evaluation board to the Kintex-7 architecture, replacing incompatible primitives, restructuring clock distribution, and manually configuring chip initialization in place of an unsupported soft-processor-based approach. Functional verification of the ADC and DAC channels, followed by closed-loop testing combining both converters with real-time filtering, confirmed correct operation of the integrated system. These results establish a working hardware and firmware baseline for the CAPTAN+X platform, positioning it for future inclusion in the laboratory's growing MBSE modeling effort.

Espinoza, David [Illinois U., Urbana (main)]↗