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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 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Study of guided wave transmission through complex junction in sodium cooled reactor

Ultrasonic guided wave techniques are seen as suitable candidates for the inspection of welded structures within sodium cooled fast reactors (SFR), as the long range propagation of guided waves without amplitude attenuation can overcome the accessibility problem due to the liquid sodium. In the context of the development of the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID), the French Atomic Commission (CEA) investigates non-destructive testing techniques based on guided wave propagation. In this work, guided wave NDT methods are applied to control the integrity of welds located in a junction-type structure welded to the main vessel. The method presented in this paper is based on the analysis of scattering matrices peculiar to each expected defect, and takes advantage of the multi-modal and dispersive characteristics of guided wave generation. In a simulation study, an algorithm developed using the CIVA software is presented. It permits selecting appropriate incident modes to optimize detection and identification of expected flawed configurations. In the second part of this paper, experimental results corresponding to a first validation step of the simulation results are presented. The goal of the experiments is to estimate the effectiveness of the incident mode selection in plates. The results show good agreement between experience and simulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Applicability of HBU Fuel Cask R and D Project to Al Clad Fuel Storage - 18558

Aluminum clad spent nuclear fuel (SNF) not destined to be reprocessed at Savannah River's H Canyon will eventually be moved from existing wet and dry storage facilities into standardized dry storage systems (e.g., Universal Canisters), ideally suited for disposal. With the current lack of disposal capacity in the U.S., placement of this SNF into dry storage requires plans for demonstrating to the regulator the safety of both the extended storage and the subsequent transportation to a future disposal facility of this SNF. For commercial SNF, concerns related to extended dry storage and subsequent transportation have been, and continue to be, addressed through several demonstration projects including, but not limited to, one for the lower burnup SNF at Idaho National Laboratory (INL) and one for the higher burnup SNF at Dominion's North Anna Power Station. Lessons learned from the more recent Orano led higher burnup demonstration project (HBU Demo Project) provides relevant information for advancing the concept of a 'smart cask' system for dry interim storage, which will allow for collection of data to support the safety demonstration for extended dry storage and subsequent transportation of the aluminum clad SNF. This 'smart cask' would eliminate the need for multiple demonstration projects that may have to be performed for different types of aluminum clad SNF designs and/or for potentially age distinguishing parameters, such as burnup or initial enrichment. This 'smart cask' would be developed and deployed by leveraging the HBU Demo Project's experience associated with modifying and deploying a TN-32B bolted carbon steel vertical storage cask system to monitor key parameters related to the criticality, structural, confinement, and thermal safety of the SNF and the canister system containing this SNF. This paper will focus on how the extensive design, licensing, and fabrication experience accumulated by the HBU Demo Project's cask system can be leveraged to support the design, licensing, and fabrication of a 'smart cask' system intended to support/confirm the interim dry storage, potential extended storage, and subsequent transportation of aluminum clad SNF, such as the Advanced Test Reactor (ATR) SNF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

TEM Characterization of U-Mo Irradiated with High-Energy Xe Ions

U-Mo alloys (with 7-10 wt.% Mo) with a low U-235 enrichment (< 20%) have been developed to be used in high-performance research and test reactors, replacing the current highly enriched uranium (HEU) fuels. A number of in-pile irradiation tests of U-Mo fuels show that drastic microstructural changes occurred during reactor irradiation, including the formation of U-Mo-Al interaction products and large gas-filled pores. These microstructural changes might lead to breakaway swelling and deterioration of plate mechanical integrity. Hence, it is critical to understand the underlying mechanisms of these microstructural changes in order to improve the irradiation behavior of UMo fuels. Heavy ion irradiation was employed in this study to investigate separate effects of the irradiation behavior of U-Mo fuels. 1.7 mm diameter disks punched out from UMo/ Al dispersion fuel mini-plates were irradiated with 84 MeV Xe ions at the ATLAS accelerator at Argonne National Laboratory (ANL). This particular ion source was selected for two reasons: (1) to simulate the irradiation response of the fuel to fission fragment bombardments, and (2) to study fission gas behavior, as fission gas bubble formation and growth is the primary cause of U-Mo fuel swelling. The ion-irradiated samples have been characterized with scanning electron microscopy (SEM). This study focuses on the characterization of the samples using transmission electron microscopy (TEM). With the high magnification details, microstructures in U-Mo along the ion penetration depth were revealed. The U-7Mo/Al dispersion fuel sample irradiated with 84 MeV Xe ions to a dose of 2.9 x 10{sup 17} ions/cm{sup 3} was characterized using TEM. The TEM results exhibit that the sample microstructure varies along the ion penetration depth. Near the sample surface (zone A) is a layer of nano-crystalline U-Mo-Al interdiffusion product. After passing through zone A, the Xe ions resided in zone B, where gas bubbles formed in U-Mo. TEM images of this region displayed that large gas bubbles formed on grain boundaries and high-density nano-bubbles lay inside grains. The random distribution of intragranular bubbles indicates that the irradiation conditions of this sample cannot lead to gas bubble superlattice formation. Outside of the Xe penetration depth (zone C), significant amount of α-uranium precipitates were identified. The fact that no α-uranium precipitates were found in zone B suggests that the dose rate in this irradiation is sufficient to facilitate a phase reversal by means of a displacement spike mechanism. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Design of a Loss of Coolant Blowdown Capsule for Remote Assembly with High Burnup Fuel

The Transient Water Irradiation System is an enhanced capability capsule type irradiation vehicle designed to support fuel safety research for light water reactor specimens in the Transient Reactor Test Facility and is designed to simulate loss of coolant and reactivity-initiated accidents. The capsule was designed, deployed, and commissioned with fresh fuel specimens to validate instrumentation and prepare for previously irradiated high burnup specimens. The irradiation system features an extensive in-situ instrumentation package to detect phenomena typical to light water reactor fuels. To accommodate the assembly with high burnup specimens inside the Hot Fuel Examination Facility, and to ensure instrumentation integrity is maintained throughout assembly, the design was updated to support remote handling. The updated design features a hinge mechanism which allows for remote pre-irradiated specimen loading and assembly while protecting sensitive instrumentation by relocating during loading. Fixtures and equipment have been developed to handle the experiment capsule and components in cell, and to support remote assembly. The experiment module also includes radiation shielding and contamination control to support operations after removal from the hot cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Consulting on HTGR Design and Analysis Methods - Lommers

The Advanced Reactor Technologies – Gas Cooled Reactor (ART-GCR) program has undertaken an effort to develop point designs of a gas-cooled, graphite moderated irradiation test reactor (TR) and a larger demonstration reactor (DR). The Department of Energy (DOE) is specifying requirements and performance metrics for these two systems. Both designs pose challenges, and their development will benefit from interaction with a potential vendor. The work defined in this subcontract supports the effort to generate point designs for gas-cooled test and demonstration reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Calculation of Neutron Damage to Sodium Coolant Inlet Plenum

The coolant inlet plenum is a life limiting component of a Sodium-cooled Fast Reactor (SFR). The inlet plenum is a fixed reactor component which receives sodium from the cold side of the sodium pool redan and directs it into the inlet nozzles of the fuel assemblies. Neutron leakage from the reactor core causes irradiation damage to the steel components of the inlet plenum leading to a reduction in strength. The life of the reactor is essentially tied to the acceptable strength degradation of the inlet plenum. Therefore, typical SFR designs employ a neutron reflector/shield as an integral feature of the fuel assembly. The role of the axial reflector/shield is to: 1) direct fast neutrons back to the core, and 2) attenuate the neutron flux to minimize neutron interaction with the inlet plenum. The longer the reflector and the greater the share of steel within the reflector design, the greater the attenuation. The axial reflector is heated by gamma rays from the core, and must have sufficient coolant penetrations to ensure adequate cool-ability. Also, the sodium coolant must pass through the reflector before reaching the fuel rods above it. The present work investigates varying reflector geometries in support of the Versatile Test Reactor (VTR), a SFR research reactor planned to restart fast reactor irradiation capabilities in the U.S. DOE laboratory complex. This work will demonstrate that the neutron damage to the inlet plenum is relatively insensitive to the number and arrangements of coolant channel penetrations through the reflector.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

ATF-2 Sensitivity M&C 2023 (Slides)

Irradiation experiments in the Idaho National Laboratory's Advanced Test Reactor are typically assumed to have little effect on one another. This assumption does not hold true for certain experiments in close proximity. To evaluate the impacts on safety and programmatic parameters of experiments in the center flux trap, the contents of the adjacent H and inner-A positions are modeled with a range of possible irradiation targets. First, neutron flux maps with experiments in those positions are compared to those of a water-filled configuration. Then, several safety and programmatic parameters for a generic accident-tolerant fuel test train are calculated. It is shown that these parameters can exhibit considerable sensitivity to the contents of the H and A experiment positions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Neutron characterization of UN/U-Si accident tolerant fuel prior to irradiation

Neutron imaging and tomographic reconstructions provide non-destructive three-dimensional characterization of nuclear fuel pellets that are typically about a cubic centimeter in volume with energy-resolved neutron imaging at a pulsed neutron source offering unique characterization not possible at a reactor source. Recent measurements at the Los Alamos Neutrons Science Center (LANSCE) studied a series of uranium nitride / uranium silicide ceramic composite pellets. This paper provides examples of spatially resolved measurements of cracks, isotopic enrichment and density with a (current) spatial resolution of approximately 75 microns. Phase and texture measurements were also performed but are not discussed in this paper. The UN/U-Si systems are under consideration as candidate systems for accident tolerant fuel. The examples are taken from measurements on a total of 9 pellets of differing UN/U-Si formulations (2 enrichment levels), U{sub 3}Si{sub 5} reference formulations (two enrichment levels) and pellets with known flaws. The {sup 235}U enrichments ranged from 0.2 to 8.8 wt.%. Pellets of the same composition are scheduled for irradiation in the Advanced Test Reactor this year. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Precipitation of Epsilon Particles in Metal Doped Ceria Irradiated with Energetic Ions

Metallic particles precipitated from fission products in oxide fuel matrix (e.g., UO{sub 2}) have a great impact on the thermal and mechanical properties of the fuel. A burn-up of 40 MWd/kg U results in the conversion of 4% of the uranium to approximately 3% fission products and 1% transuranium elements. Epsilon phase is the major metallic phase of the precipitates with a hexagonal-closed packed (HCP) crystal structure. It can exist in a variety of metallic forms in UO{sub 2}, but typically, it consists of five metals: Mo, Pd, Rh, Ru and Tc. Epsilon particles can form over a wide range of stoichiometries in the five metal system. Simulated compositions with Re as a surrogate for Tc have been investigated as a waste form for {sup 99}Tc, an isotope with a long half-life of 2.13 10{sup 5} years and a high solubility under oxidizing conditions in addition to Mo in the corrosion process. There is a pressing need to understand the factors influencing the formation and growth of the metallic phase in fuel. Efforts are currently underway to perform careful neutron irradiation of nuclear fuel under controlled conditions in a test reactor, recover metallic phases that may be present, and characterize the resulting structure and chemistry. Due to the long duration, high cost and radiological concerns associated with neutron irradiation and sample characterizations, ion irradiation studies of a surrogate material with a full control of the processing parameters can emulate different stages of the particle precipitation, which can add a tremendous value to reactor-based studies. This presentation reports our recent ion irradiation results on the metallic particle precipitation under various irradiation and thermal annealing conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Hybrid Zircex Hydrochlorinator

This project was driven by the desire to create easier and faster ways to load Advanced Test Reactor (ATR) Fuel Elements into the Hybrid Zircex hydrochlorinator. The primary requirements for this project were to minimize the potential release of radioactive contamination, minimize maintenance operations to replace or remove the lid seal, and allow for remote usage. Three preliminary designs were made. These designs allow ATR fuel elements to be loaded into the hydrochlorinator more efficiently while fulfilling the requirements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Risk Analysis of Radiological Release from Pu-238 Targets During Manual Handling

Pu-238 Isotope Production Targets are routinely installed in the Advanced Test Reactor (ATR) core, transferred, and stored in the spent fuel canal. These evolutions involve manual handling and manipulation of the targets underwater using long handled tools. The ATR Safety Analysis Report (SAR) postulates a design basis accident which results in damage from manual manipulation of targets, and radiological consequences must be determined for receptors inside the reactor facility, as well as public receptors. This presentation presents the analysis used to determine the radiological consequences due to potential target damage in the ATR canal. The analysis considered radionuclide release fractions, damage ratios for handling evolutions, and entrainment of radionuclides in the canal water.

Advanced Test Reactor↗
DOE OSTI2017

Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

This paper reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. Finally, these characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Leveraging the High Flux Isotope Reactor for nuclear fuel development: a review of experiments, facilities, and capabilities

Materials testing reactors (MTRs) have been used to develop in-core nuclear fuels and materials since the outset of the nuclear power industry. However, the closure of prominent MTRs worldwide and protracted construction timelines for new facilities have increased reliance on existing infrastructure for near-term irradiation testing needs. One facility that can support these needs is the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. HFIR boasts the highest steady-state neutron flux in the Western Hemisphere and, among other roles, has been used to rapidly administer high fluences on fuels and materials for fission and fusion reactor applications. This paper reviews HFIR facilities and infrastructure, fuel-bearing irradiation experiments conducted in HFIR, and select nonfueled experiments that demonstrate advanced techniques transferable to fuels experiments. Collectively, these examples underscore HFIR's potential role as a nuclear fuels testbed supporting both the existing reactor fleet and advanced reactor fuel development.

Fuel qualification↗
DOE OSTI2022

Transformational Challenge Reactor – On the Application of Design for Additive Manufacturing (DfAM) Techniques to the Conception of Nuclear Core

Additive manufacturing (AM) technologies are radically changing the way objects are designed and manufactured. They allow building by deposition and solidification of material layer by layer, enabling the possibility to create simple and complex features alike, almost seamlessly. Generally, the design optimization process requires to define objectives, design variables and constraints. Additive manufacturing does not challenge this process per se but does allow designers to completely redefine the constraints space as the ones originating from fabrication can be considerably relaxed compared to more “traditional” manufacturing. Thus, design optimization becomes naturally far more responsive to the actual physics being solved and considerably less influenced by fabrication limitations, leading to dramatically different designs. To take advantage of these new opportunities, so-called Designing for Additive Manufacturing (DfAM) techniques are emerging. Development of design techniques specifically tailored for additive manufacturing is warranted because, considering AM, the design space is typically considerably larger than with traditional manufacturing. The ability to explore the design space efficiently is of paramount importance for designers. This study proposes to investigate and apply some of these DfAM techniques to the conception of nuclear core. The goal being to assess if these new methods can be applied to core design and if core design could benefits from additive manufacturing technologies. After a brief investigation on the pertinence of some DfAM techniques for core design, algorithms are proposed and a workflow is established to carry neutronics and steady-state thermal-hydraulics analyses. To diminish the work load, the workflow has been automated using python modules. These modules allow the rapid creation of input files, post-treatment of output files and visualization. To test the pertinence of the proposed workflow, three test cases have been investigated: a research and test reactor, a micro-reactor and a space propulsion reactor. These test cases offered a variety of objectives, constraints and operating conditions. It is observed that the proposed workflow is capable of converging quickly and efficiently to valid design solutions. It is then concluded that DfAM techniques can be applied to core design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

The High Temperature Test Reactor (HTTR) is a graphite moderated thermal reactor with a nominal power output of 30 MWth. The purpose of this work is to prepare a set of neutron cross sections for use in the transient analyses of three Loss of Forced Cooling (LOFC) events planned for this reactor. The cross sections are generated with the Serpent 2 3 D Monte Carlo code to include the axial heterogeneity and the strong axial coupling of this core. The approach yields good agreement of results when compared to previously used 2 D lattice models, which require a fine spatial discretization and, surprisingly, comes at a higher computational cost. In addition, the temperature distribution must be considered to obtain reasonable cross sections and the axial leakage component dominates the spectral effects in the core. Consequently, a temperature model as a function of the fuel and moderator temperatures is adopted, and the core is depleted to 390 effective full-power days (EFPDs). A full tabulation of cross sections is prepared at this burnup point for a variety of fuel and moderator temperatures. Next, the multiphysics reactor application MAMMOTH is used to evaluate the quality of these cross sections using a diffusion solver. As expected, the homogenization error in the cross sections is significant and the Super Homogenization (SPH) correction from MAMMOTH is necessary to preserve key reaction rates. The SPH corrected MAMMOTH results are in excellent agreement with the Serpent results and reproduce both the reference power profile and temperature coefficients at each tabulation point. This fact confirms the accuracy of the SPH corrected cross sections and demonstrates the maturity of using Serpent 2 and MAMMOTH for 3 D cross section generation, even for reactors as complex as the HTTR. Finally, the authors recommend that the cross sections should be parametrized with the local burnup and that a full core burnup calculation with coupled thermal fluids would provide a better estimate of the initial condition for any subsequent transient analysis. Unfortunately, this was not within the scope of this work, and direct analysis with these tabulations should be considered as a first order approximation, at best.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Projection Analysis for ATR Irradiation of the AFC-FAST Experiment

Analyses of the Advanced Fuels Campaign Fission Accelerated Steady-state Test (AFC-FAST) in the Advanced Test Reactor are presented. A detailed methodology was employed to better account for uncertainties in the planned power and duration of sequential reactor loading cycles. By performing coupled depletion analyses at multiple power levels and durations, the differences in experiment heating outputs can be found. The effects of these uncertainties upon multiple experiment configurations were assessed in an effort to streamline the process of planning for and documenting future irradiations. In conclusion, the data generated from this work have been used to help inform assumptions on subsequent projections to perform only a nominal case depletion.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Analysis of Radiological Release From Fueled Irradiation Experiments During Manual Handling (Slides)

Irradiation experiments are manually handled at the Advanced Test Reactor (ATR) by qualified operators using long handled tools. Standard handling evolutions include insertion and removal from the reactor vessel, transfer to/from a storage location or cask in the ATR canal, and handling/reconfigurations at a canal working tray. Such routine handling has the potential to result in mechanical damage to the experiment boundary which is credited to retain fission products. Since damage can occur due to operator error, this is an anticipated occurrence. This work determines the radiological consequences to receptors inside the ATR facility, as well as public receptors. Given the wide variety of fuel types tested at the ATR, a generic approach to the analysis is taken. The radiological inventory is determined to bound a variety of fuel types (e.g., ceramic and molten fuel matrices) and fission powers on the U-235 enrichment spectrum. The source term analysis considers different release fractions to bound different fuel types and burnups. The postulated handling events occur underwater; thus, retention of the released isotopic content is considered within the canal water column. Retention of radionuclides in the water column is determined using a modified approach of United States (US) Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.183. Radiological dose to the facility receptor is determined using a compartment release model. Dose to the public receptor is determined using atmospheric dispersion models using site specific atmospheric conditions with the use of the Radiological Safety Analysis Computer (RSAC) program version 7.2.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2019

Innovative Implementation of Consolidated Dry Fuel Storage at the INL - 19655

An innovative nuclear-fuel storage bucket design provides increased fuel storage capacity at the Irradiated Fuel Storage Facility (IFSF) on the Idaho National Laboratory (INL) Site. The IFSF is a dry storage facility that is at approximately 93% of its fuel storage capacity. This innovative fuel storage bucket provides for higher density packaging, thereby allowing for the receipt of more fuel at the IFSF and subsequent completion of the wet-to-dry fuel transfer scope at the INL's Idaho Nuclear Technology and Engineering Center (INTEC). The subject fuel originates from the Advanced Test Reactor (ATR) at the INL Site. All spent nuclear fuel (SNF) at INTEC is required by the Idaho Settlement Agreement (ISA) to be out of wet storage by December 31, 2023. The ISA is a tri-party agreement between the U.S. Department of Energy, the U.S. Navy, and the State of Idaho. Currently, the INTEC CPP-666 pools contain all of the INL's SNF remaining in wet storage, and INTEC's Spent Fuel Project is on schedule to meet the ISA deadline. Interestingly, the INL Site's environmental cleanup contract awarded to Fluor Idaho in 2016 also initially required the Spent Fuel Project to continue SNF receipts from the ATR into wet storage at the CPP-666 pools. The continued receipts of ATR SNF into the CPP-666 basin were able to be redirected to the IFSF because of the new storage package design, enabling the Spent Fuel Project to meet the ISA requirement to have all SNF out of wet storage by 2023. Complicating the need to redirect ATR receipts from wet storage is the lack of interim storage capacity at the ATR site. This lack of ATR storage capacity threatened to terminate reactor operations if the SNF could not be received at INTEC. The contractual requirement for continued ATR SNF receipts to prevent termination of reactor operations is resolved by the implementation of the innovative consolidated dry storage configuration in the IFSF. Consolidated fuel storage at the IFSF provides increased fuel storage capacity to facilitate meeting the ISA and the Fluor Idaho contract requirements. It provides for the completion of the ATR wet-to-dry fuel transfer scope and supports continued operation of the ATR by allowing the receipt of additional ATR SNF directly into IFSF dry storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗