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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Software Quality Assurance for EBR-II Fuels Irradiation and Physics Database (FIPD)

The Fuels Irradiation and Physics Database (FIPD) is an ongoing DOE project on archival of the EBR-II metal-alloy fuel irradiation experiments. As part of its use in support of license applications, the Quality Assurance Program Plan (QAPP) was drafted and endorsed by NRC in an effort to demonstrate its compliance with regulatory expectations. Software Quality Assurance (SQA) for the physics portion of FIPD is intended to qualify the calculated quantities such as fuel and cladding temperatures, neutron fluence and axially varying burnup estimates for irradiated fuel elements. This report covers the initial evaluation of SQA status of three neutron physics and thermo-fluid codes (REBUS, RCT and SE2RCT) that form the basis of calculated quantities for as-irradiated characteristics of the tested metallic fuel elements. The report also introduces an SQA plan to address the identified deficiencies. The REBUS, RCT, and SE2RCT codes are all part of the Argonne Reactor Code (ARC) code system. There is considerable knowledge and experience on REBUS and RCT but relatively less on SE2RCT. During FY2021, efforts focused on an assessment of how the data in the EBR-II Physics and Analysis DataBase (PADB) is generated with SE2RCT and used in FIPD. Additional tasks included considerations of uncertainties for power estimates in REBUS and RCT calculations and their impact on the combined RCT methodology. The RCT software usage in FIPD was assessed this year and the input/output details studied. A “requirements” document was created that identifies the key features of the RCT software being used in FIPD that need to have SQA documentation. A brief discussion on the history of RCT and its input is included in this report along with the basic SQA roadmap laid out in the requirements document. The SE2RCT software usage in FIPD is still being studied noting that there is no current manual. As part of the work done this year, two bugs were identified in the SE2RCT software which have a minor impact on the accuracy of the results it produces. No requirements document has been created, but one identified feature of SE2RCT being used that needs verification was its fuel pin temperature calculation. The work completed this year confirms that the approximations which will be included in the software verification report for SE2RCT are accurate. In addition to software quality assurance work for RCT and SE2RCT, an automated verification framework is proposed to simplify the software quality assurance process. The purpose of this framework is to streamline code verification and documentation while minimizing repetitive tasks for code developers and reviewers. The reduction of repeated input (between reference solution, software, and documentation input) throughout the SQA process reduces potential for human errors during the preparation of the supporting software quality records. The automation of the verification and documentation process proposed for this project leverages the existing verification structure already in place for the SAS4A/SASSYS-1 code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Status Report on Development of a Cask to Enable Pulsed Neutron Characterization of Irradiated Fuel

We present the design of a shielding cask that allows pulsed neutron characterization of irradiated fuel rodlets prior to their destructive examination in hot cells. The goal is to provide complementary and informative measurements that will inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm3 to 1 cm3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging. The capability will identify regions of atypical behavior and provide cost effective bulk characterization of entire samples prior to destructive evaluation in hot cells. This effort is a collaboration between LANL, INL, ORNL, and UC Berkeley. A key facet of the initiative is the ability to cost effectively ship irradiated fuels from INL to the pulsed spallation neutron source at LANSCE. The irradiated fuel would be loaded in the custom designed cask (designated RaMHaM) at INL hot cells which then could be shipped in a BRR Type B shipment container between INL and LANL. No hot cell is required after the loading of the sample at INL in the pathway laid out in this report, greatly simplifying handling at LANL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Nuclear fuel irradiation testbed for nuclear security applications

The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO 2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO 2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Characterization of Pinhole Collimators for High-Resolution Gamma Imaging of Irradiated Fuel

Post-irradiation examination (PIE) of nuclear fuels requires imaging tools capable of resolving isotopic and spatial features with high throughput. This project contributes to a proof-of-concept effort aimed at advancing gamma emission tomography (GET) by evaluating novel fine-aperture pinhole collimators. Two Rose’s metal collimators, 100 µm (20° acceptance angle) and 350 µm (30° acceptance angle), were prototyped and characterized for their effectiveness in transporting gamma rays through the pinhole aperture. To support data collection, a Python-based data acquisition system was developed to coordinate a rotation stage, linear stage, and CZT detector, reducing latency in high-rate gamma event logging to one second per acquisition. Queue-based file writing enabled seamless real-time data capture for count rates up to 35,000 counts per second (cps). List-mode parsing algorithms were implemented to differentiate single and simultaneous gamma interactions for future tomographic reconstruction. Detector response was evaluated in both spectroscopy and list mode acquisition methods across varying source distances to confirm absolute and collimator efficiencies. Preliminary efficiency figures suggest effective collimation of gamma-rays with energies below 700 keV, with ~4% residual intensity through the aperture for Cs-137. The impact of collimator geometry on image quality is currently being evaluated. This groundwork supports the ongoing development of a sub mm resolution cone-beam CT system for imaging fuel phantoms, an essential step toward improving GET efficiency and accelerating nuclear fuel qualification efforts.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2023

Draft Summary Report on Irradiated Fuel Handling and Management for LOTUS

The National Reactor Innovation Center (NRIC) has conceptualized the design of the Laboratory for Operation and Testing in the United States (LOTUS) test bed to provide the United States Department of Energy (DOE) with the infrastructure necessary to make advanced reactor designs available for commercial developers. LOTUS will provide a test bed to developers with the capabilities of supporting a wide range of experiment design possibilities. Upon completion of the developers’ operations and experiments within the test bed, the irradiated fuel, reactor components, and other experiment materials must be removed. Idaho National Laboratory (INL) possesses significant capabilities for radioactive material handling such as casks, carts, and forklifts. However, given the unique environment presented by the NRIC-LOTUS test bed, located inside the Zero Power Physics Reactor (ZPPR) Perimeter Intrusion Detection and Assessment System (PIDAS) area at the Materials and Fuels Complex (MFC) and the complexity of novel removal activities of recently operated reactor experiments through the new proposed access tunnel. The efficacy was not apparent for existing equipment to provide all the needed capability. To bridge the potential gaps in cask designs, storage, and transportation, NRIC requested the development of trade studies for the transfer, handling, and storage requirements of irradiated fuel salts and other radioactive materials. NRIC directed Boston Government Services, LLC (BGS) to perform the trade studies and develop a report analyzing alternatives. In addition to the BGS reports, the Idaho National Lab’s (INL), provided by the first potential user’s Advanced Reactor Development (ARD) team, prepared a feasibility study for the storage of specific irradiated fuel within the existing ZPPR vault.This summary report is intended to present the trade studies, options, and alternatives that were investigated. The maturity level of LOTUS, the reactor, fuel salt containers, gloveboxes, and reactor testing campaign and concept of operations were not at a level sufficient to base critical decisions on. This report is not intended to present a final recommendation. The final recommendations for fuel storage location, transport, handling equipment, and operations will be made in FY 2024 and will be based on known materials, test campaign requirements, funding, and final analysis of the fuel and equipment to be used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Experiment Assembly and Testing in Hot Cell in Support of Advanced Fuels Irradiation Testing

Follow-on experimentation of irradiated nuclear fuel is critical to understanding fuel behavior at different stages of the fuel lifecycle. Transient experiments using fuel with high burnup can enlarge understanding of fuel fragmentation, relocation and dispersal under various reactor accident situations. Assembling experiments with such fuel requires remote operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Transmission Electron Microscopy based Characterization of a U-20Pu-10Zr Fuel Irradiated in Experimental Breeder Reactor-II

U-Pu-Zr metallic fuels are important fuel candidate for future advanced and/or test reactors. To better understand fuel performance and guide future fuel development, a U-20Pu-10Zr (in weight) metallic fuel irradiated in the Experimental Breeder Reactor was revisited using advanced electron microscopies. This fuel was irradiated to a burnup of 7.6 % at a peak cladding temperature of 521 °C. In this research, several transmission electron microscopy samples were extracted from different but representative radial locations of the fuel cross section using focused ion beam technique. Phase identification and chemical analysis with sub-micron spatial resolution was carried out using scanning transmission electron microscopy. Multiple phenomena that are critical to fuel performance, such as Zr phase and its distribution in a different matrix, were revealed in an unpreceded and highly detailed manner. An improved understanding of fuel reconstruction in U-Pu-Zr metallic fuel under reactor irradiation is provided.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Sample Preparation for 3D Characterization of Irradiated Fuel

Thorough characterization of nuclear fuel is essential when striving to understand its microstructural evolution in response to varying irradiation conditions. A detailed understanding of the microstructural evolution of irradiated fuel can be useful when predicting the fuel’s response to off-normal conditions and related phenomena such as high burnup fuel fragmentation (HBFF) and fission gas release. This sort of characterization and the study of complex phenomena like HBFF that are related to fuel performance are an important part of the US Department of Energy Office of Nuclear Energy’s Advanced Fuels Campaign (AFC). Multiple characterization techniques are commonly used to perform post-irradiation examination (PIE) analysis on an irradiated fuel sample, but many of these techniques are limited to the 2D surface analysis of the fuel pellet. Although there is a lot of useful information to be gained by analyzing the fuel pellet surface, a 2D analysis provides only an approximation of the nature of the material’s features, such as porosity and grain structure. Therefore, it is critical to analyze the 3D structure of the material, as materials exist and respond to conditions in all three dimensions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Preconceptual Design of Irradiated Fuel Salt Management System

The National Reactor Innovation Center was established by the U.S. Department of Energy to accelerate the demonstration and deployment of advanced nuclear reactors. To meet this mission, the National Reactor Innovation Center has developed the Laboratory for Operations and Testing in the U.S. (LOTUS) to support the first fast-spectrum molten salt reactor demonstration. The goal of this research is to understand the system requirements that may be applied to a molten salt reactor experiment in the LOTUS test bed and to perform systems analyses and preconceptual development of a storage container for the management of the irradiated molten salt fuel. This study proposes a set of irradiated fuel salt management processes for the molten salt reactor, from defueling to storage or disposal. For that, the historic Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has served as a crucial source of data for a mock application. This paper proposes a potential geometry for storing irradiated fuel salt based on MSRE data and the Monte Carlo N-Particle radiation transport code. The geometry was analyzed for criticality safety under abnormal situations, such as water ingress and interaction between containers. Based on the determined configuration of the storage system, the decay heat produced from the irradiated fuel salts was calculated to provide a reasonable time scale of the dormancy period. Using the same geometry assumed above, the neutron and gamma dose rates from the irradiated fuel salt at the end of the dormancy period were estimated, which will be useful for understanding the handling and radiation protection requirements for handling the irradiated fuel salt. Lastly, a reactor in the LOTUS test bed could use chloride-based fuel as opposed to the fluoride-based fuel used in MSRE. Thus, a literature study provided data to understand the similarities and differences between the characteristics of fluoride- and chloride-based salts in terms of salt chemistry, material characteristics, postirradiation behavior, neutronics, thermal hydraulics, and heat removal. This study, together with the proposed irradiated fuel salt management processes, preliminary criticality calculations, decay heat, and dose rate estimates, provides a strong basis for future investigation into the management of irradiated fuel salts generated during molten salt reactor technology demonstrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

USABILITY EXTENSION OF THE ŠKODA VPVR/M CASK FOR TRANSPORTING IRRADIATED FUEL ASSEMBLIES

New abstract from the final version being submitted now: This paper discusses the evolution and enhanced usability of the ŠKODA VPVR/M cask for the transport of irradiated fuel assemblies, particularly within the context of the demand for the delivery of newly appearing irradiated HEU fuel types for which the cask did not yet have a license to transport. Over time, the cask’s internal basket construction has demonstrated notable adaptability to accommodate various exotic HEU fuel types from research reactors of differing origins. The paper outlines sever-al custom internal baskets developed for specific fuel types, including those from Belarus and Serbia, as well as from Georgia, Uzbekistan, and MNSR cores, as well as a recently designed basket for MTR and TRIGA assemblies. The findings high-light the high flexibility and adaptability of the cask, supported by successful rede-signs and licensing efforts, underscoring its value for the safe and secure transport of nuclear material. Old abstract from the draft version that was already approved: The Russian Research Reactor Fuel Return (RRRFR) program, since its inception, has continuously used the ŠKODA VPVR/M Cask fleet designed for the repatriation of irradiated highly enriched uranium (HEU) fuel. As the program progressed (from shut-down and a quasi-abandoned reactor, and/or as it began to include fuels of Chinese and US origin), new challenges emerged for the transport Cask. These were fuel types that had not yet been licensed for the Cask. Although these requirements did not arise during the design of the basic ŠKODA VPVR/M Cask, as revealed by the retrospective analyses, the Internal Basket of the ŠKODA VPVR/M Cask gives a high degree of flexibility to accommodate additional fuel types. This paper provides a brief overview of the ŠKODA VPVR/M Cask, which holds a B(U) type license, and introduces the different types of Internal Baskets that have already been licensed to transport so-called exotic irradiated HEU fuel types, in addition to the original license. The paper presents a new Internal Basket design for accommodating MTR-type and TRIGA-type irradiated HEU fuel assemblies. This includes a detailed presentation of the design basis and the new MTR-TRIGA Internal Basket, as well as the licensing matters of the package under the name ŠKODA MTR-TRIGA Cask, and the conformity test (dry- and wet-run) operations made to verify compliance with the new Internal Basket. Then, as a summary, the usage record for the Cask fleet is presented, and finally, the paper concludes with the consolidated experiences gained during the utilization of the ŠKODA VPVR/M Cask fleet, emphasizing the high degree of Cask flexibility ensured by the Internal Basket’s construction.

42 - ENGINEERING↗
DOE OSTI2022

An investigation of the failure modes in U-10Mo monolithic fuel irradiated to high burnup

Here, this study investigated possible failure mode sequences in U-10Mo monolithic fuel irradiated to very-high-burnup. The U-Mo fuel plate used in this study was characterized using scanning electron microscopy (SEM), wavelength dispersive spectroscopy (WDS), and image analysis techniques to investigate how microstructural features may initiate crack formation and propagation through the microstructure, potentially leading to blistering. Distinctly large fission gas pores (FGPs) were observed to preferentially align 5-15µm away from the U-10Mo/Zr interface, parallel to the Zr diffusion barrier. Preferential growth, alignment, and interconnection of FGP could initiate blistering in the fuel plate by creating a fission gas channel, parallel to the U-Mo/Zr interface. FGP alignment in the fuel phase near the U-Mo/Zr interface is believed to be one of the precursors to crack and Type 2 blister formation in monolithic U-Mo fuels. The chemical maps revealed differences in fission product behavior near the U-Mo/Zr interface such that Nd remains in the fuel matrix at the FGP sites. On the other hand, other fission products like Xe and Cs can get trapped in FGPs by Nd in the fuel phase and/or diffuse toward the Zr diffusion barrier.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

FIPD: The SFR metallic fuels irradiation & physics database

The DOE Advanced Reactor Technology (ART) program has supported efforts to recover and preserve metallic fuel data generated throughout the US sodium-cooled fast reactor (SFR) program. Those efforts have been focused on establishing databases of the experimental data that were mainly generated during the Integral Fast Reactor (IFR) program including data generated at Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and Transient Reactor Test Facility (TREAT) reactors, as well as out of pile data. The data is essential for future licensing activities of metallic fuel based advanced fast reactors. This paper describes the development of the SFR Metallic Fuels Irradiation & Physics Database (FIPD) and covers the scientific knowledge available in the database. Furthermore, the architecture of the FIPD is described by showing the available reactor operation data, fabrication, and post-irradiation examination (PIE) data, and other documents. The applications of the fuel database are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Preliminary Evaluation of Loading DOE Standard Canisters in the INL CPP-603 Irradiated Fuel Storage Facility - 20543

This paper looks at the equipment and operations necessary to load United States Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) into DOE Standard Canisters in the CPP-603 Irradiated Fuel Storage Facility (IFSF) in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The fuel ready for loading would come from fuel storage canisters in the CPP-603 facility. The paper describes the facility, the fuel types, the DOE Standard Canisters, and existing equipment; lists the needed loading operations; reviews facility features and equipment to perform the operations; and then lists the decisions, analyses, designs, demonstrations, and modifications that will be needed to perform the loading of DOE-owned SNF into DOE Standard Canisters in the CPP-603 IFSF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Elucidating the effect of minor-actinide addition on fuel-cladding chemical interaction in an HT-9 clad U-Pu-Zr metallic fuel irradiated to 6.15 at.% burnup in EBR-II

Scanning and transmission electron microscopy (S/TEM) were used to characterize the local fuel-cladding chemical interaction (FCCI) in one cross-section taken from a HT-9 clad U-20.3Pu-10Zr-1.2Am-1.3Np (in wt.%) fuel irradiated to 6.15 at.% burnup with inner cladding temperatures ranging between 460–490 °C. Results showed that the total interaction thickness between fuel and cladding was <10 µm. Fe infiltrated the fuel to form U-Zr-Fe phases while fuel elements or lanthanides did not infiltrate into the cladding. Np was not involved in the formation of any phases in the examined locations; however, Am played a role by forming a ∼2 µm thick homogeneous Fe-Pu-Am planar front at the inner cladding wall. An oxidized Na layer existed in the fuel-cladding gap with Fe and lanthanide particles dispersed within, suggesting Na could facilitate the transport of fuel and cladding constituents. Secondary phases, including an FCC Zr-rich phase, lanthanide phases, and α’-Cr(Fe) were identified in the outer fuel and FCCI regions. Furthermore, this study suggests that, for the irradiation conditions specific to this cross-section, minor actinides have little impact on FCCI behavior beyond what would be observed in typical HT-9 clad U-Pu-Zr fuel pins systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

3D pore structure reconstruction of fluff region in top of EBR-II irradiated fuel by X-ray micro-computed tomography

Metallic fuels comprised of U-Zr binary and U-Pu-Zr ternary alloys are a promising fuel candidate for sodium-cooled fast reactors and possess many advantages over ceramic fuels such as higher uranium density and improved thermal conductivity. Fluff is a highly porous structure forming at the top of metallic fuel slugs during irradiation and was extensively observed in past sodium-cooled fast reactor (SFR) experiments such as Idaho National Laboratory’s Experimental Breeder Reactor II (EBR-II). Previous investigations have shown that fluff is rich in fissile atoms and hence could have neutronics implications, but its formation mechanisms are not understood. In this work, two cube-shaped lift outs were taken from the fluff and bulk fuel regions of a high burnup U-19Pu-10Zr composition EBR-II fuel slug and characterized by micro computed tomography (Micro-CT) and energy dispersive x-ray spectroscopy (EDS). This was used to characterize differences in microscale porosity and composition between fluff and bulk fuel, and to support or oppose postulated fluff formation mechanisms. Surface EDS showed minimal differences between the elemental compositions of the fluff and fuel cubes. However, three-dimensional (3D) pore analysis from the tomography scans showed significant differences between the fluff and fuel cube pore size distributions at the microscale level. The high porosity and irregular pore distribution seen in macroscale fluff was also present within the micro-scale fluff cube, having a much higher porosity than the fuel cube which was concentrated in a few large pore networks. In conclusion, these results indicate support for a creep driven mechanism of fluff formation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Fuels Irradiation and Physics Database (FIPD) Development and Data Qualification FY24 Updates

The DOE-NE Advanced Reactor Technologies program has supported efforts to recover and preserve metallic fuel data generated throughout the past fast reactor R&D programs. Those efforts are currently focused on establishing databases for information from the experiments conducted during the Integral Fast Reactor program including data generated at EBR-II, FFTF, and TREAT reactors, as well as out of pile transient testing data. These databases are essential for future-licensing activities of metallic fuel based advanced fast reactors. This report describes the Metallic Fuels Irradiation & Physics Database (FIPD) development, content updates, data qualification status, and user support in FY24. Future plans are also provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗