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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Post-irradiation Examination and Fission Product Inventory Analysis of AGR-1 Irradiation Capsules

The AGR-1 experiment was the first in a series of Advanced Gas Reactor (AGR) experiments designed to test TRISO fuel under High Temperature Gas Reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL’s Materials and Fuels Complex (MFC). The inventory and distribution of fission products, especially Ag-110m, was assessed and analyzed for all the components of the AGR-1 capsules. This data should help inform the study of fission product migration in coated particle fuel. Gamma spectrometry was used to measure the activity of various different fission products in the different components of the AGR-1 test train. Each capsule contained: 12 fuel compacts, a graphite holder that kept the fuel compacts in place, graphite spacers that were above and below the graphite holders and fuel compacts, gas lines through which a helium neon gas mixture flowed in and out of each capsule, and the stainless steel shell that contained the experiment. Gamma spectrometry results and the experimental techniques used to capture these results will be presented for all the capsule components. The components were assayed to determine the total activity of different fission products present in or on them. These totals are compared to the total expected activity of a particular fission product in the capsule based on predictions from physics simulation. Based on this metric, a significant fraction of the Ag-110m was detected outside the fuel compacts, but the amount varied highly between the 6 capsules. Very small fractions of Cs-137 (<2E-5), Cs-134 (<1e-5), and Eu-154 (<4e-4) were detected outside of the fuel compacts. Additionally, the distribution of select fission products in some of the components including the fuel compacts and the graphite holders were measured and will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Compaction Scale Up and Optimization of Cylindrical Fuel Compacts for the Next Generation Nuclear Plant

Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of designed experiments have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel. Results from these experiments are included. The scale-up effort is nearing completion with the process installed and operational using nuclear fuel materials. The process is being certified for manufacture of qualification test fuel compacts for the AGR-5/6/7 experiment at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

A risk-informed approach to safety margins analysis

The Risk Informed Safety Margins Characterization (RISMC) Pathway is a systematic approach developed to characterize and quantify safety margins of nuclear power plant structures, systems and components. The model has been tested on the Advanced Test Reactor (ATR) at Idaho National Lab.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

Effects of thermal treatment on the co-rolled U-Mo fuel foils

A monolithic fuel type is being developed to convert US high performance research and test reactors such as Advanced Test Reactor (ATR) at Idaho National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU). The interaction between the cladding and the U-Mo fuel meat during fuel fabrication and irradiation is known to have negative impacts on fuel performance, such as mechanical integrity and dimensional stability. In order to eliminate/minimize the direct interaction between cladding and fuel meat, a thin zirconium diffusion barrier was introduced between the cladding and U-Mo fuel meat through a co-rolling process. A complex interface between the zirconium and U-Mo was developed during the co-rolling process. A predictable interface between zirconium and U-Mo is critical to achieve good fuel performance since the interfaces can be the weakest link in the monolithic fuel system. A post co-rolling annealing treatment is expected to create a well-controlled interface between zirconium and U-Mo. A systematic study utilizing post co-rolling annealing treatment has been carried out. Based on microscopy results, the impacts of the annealing treatment on the interface between zirconium and U-Mo will be presented and an optima annealing treatment schedule will be suggested. The effects of the annealing treatment on the fuel performance will also be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

Accident Tolerant Fuels↗
DOE OSTI2016

Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhance heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.

Terrani, Kurt A↗
DOE OSTI2015

Development of Metallic Fuels for Actinide Transmutation

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10 and 60%. In general, the performance of all of these substantially disparate metallic fuel alloys has been observed to be excellent, and their irradiation behaviors are generally consistent with historic norms for metallic fuels without minor actinide additions and having lower Pu or Zr contents. Future work is being undertaken with a view toward increasing the burnup potential of metallic fuels even more. Design innovations under investigation include: 1) lowering the fuel smear density in order to accommodate more swelling, 2) annular fuel geometry to eliminate the need for a sodium bond, 3) minor alloy additions to stabilize lanthanide fission products inside the fuel and prevent their transport to the cladding where they can participate in fuel-cladding chemical interaction (FCCI), and 4) coatings/liners on the cladding inner surface to mitigate FCCI and enable higher temperature operation. This paper will present the current state of development of metallic fuels for actinide transmutation in the US. Highlights will include recent results from metallic fuel casting experiments, experiments to identify alloy additions to immobilize lanthanide fission products, and postirradiation examinations of annular metallic fuels at low burnup.

actinide transmutation↗
DOE OSTI2017

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

On the Characterization of a Drop-in Miniplate Experiment–Hydraulic Analysis

The Miniplate-1, Large B (MP-1LB) vehicle is intended for future in-pile application within the Advanced Test Reactor (ATR), located at the Idaho National Laboratory (INL). In order to support the safety bases for design considerations associated with this in-pile experiment, a campaign of out-of-pile testing is presently underway at the Oregon State University (OSU) Hydro-Mechanical Fuel Test Facility (HMFTF). The purpose of out-of-pile flow testing of representative in-pile geometry is intended to provide empirical information that can provide insight regarding characteristics of the vehicle and its internals that are intended to be placed in the ATR. The objectives of this testing campaign are to 1. verify that the integrity of experimental hardware and its configuration is not compromised under anticipated in-pile hydraulic loads; 2. provide experimental data to a. quantify hydraulic characteristics of the experiment, which may be used to update experiment calculations; b. quantify the transient response of experiment hardware to expected hydraulic loads and boundary conditions; and 3. potentially identify phenomena that were not identified during design and calculation process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Physics Validation Measurements for Highly-Reactive Experiment Packages in the INL Advanced Test Reactor

The Korea Atomic Energy Research Institute is currently in the process of qualifying the Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element will be irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

NEW MATRIX COEFFICIENTS FOR THE LOBE POWER CALCULATION AND INDICATION SYSTEM IN THE ADVANCED TEST REACTOR

Theoretical coefficients are tuned for use in the Advanced Test Reactor (ATR) Lobe Power Calculation and Indication System (LPCIS). The LPCIS indicates localized fission powers in the ATR, used primarily by operators to provide irradiation of various experiments at requested power levels. The LPCIS solves an overdetermined matrix, such that the best-fit solution can be obtained from at least five of up to eleven independent indications selected for input. Hourly data logs are reviewed for three and one-half years (from November 2013 through June 2017) of ATR operation to compare localized powers indicated by LPCIS with powers indicated by an independent indication, the Water Power Calculator (WPC). WPC indications are used for the calibration of LPCIS because of assumed high reliability and theoretical simplicity. WPC is also used for automatic core protective functions, which LPCIS is not. As the WPC provides power only for four quadrants and the LPCIS indicates power in five lobes, a calculated metric is required to compare the two indications. The metric chosen for the current work has been in use for decades at ATR as a metric of LPCIS health. The proposed new coefficients better fit theory of physical phenomena involved and also provide a 40% improvement in the calculated fit.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Poster and Slide: Energy Loss: Exploring Dissipation Pathways from Ions to In-reactor

Young Researcher Association is hosting an early career poster and networking session on July 25th, 2019. The poster session is designed to provide an overview of the researcher’s experience prior to employment at INL, and their current area of research and interests. The following is a brief description of the major sections of the poster. Before starting at INL in May of this year, my research focused on exploring the effects of electronic and nuclear energy loss on the damage evolution of Ti3SiC2 MAX phase during energetic irradiations. A series of ion irradiations revealed a threshold in electronic energy loss (Se), above which the formation of a FCC-MAX structure is enhanced, accommodating point defects and alleviating lattice strain. However, systematically separating the effects reveals that increasing Se/Sn suppresses this phase formation, thus allowing the strain from point defects to increase and saturate. Presently my research supports Post Irradiation Examination (PIE) of U-Mo fuel for the USHPRR and KJRR programs. The PIE evaluation currently consists of non-destructive examination (NDE) of fuel plates irradiated in the Advanced Test Reactor (ATR), focusing on plate profilometry and fuel swelling analysis. NDE will culminate in finalizing the destructive examination (DE) test plans and execution of said DE. Additionally, I am interested in exploring the roles that different energy dissipation pathways play in the damage evolution of the U-Mo fuel, exploring the mechanisms by which the fuel remains in its meta-stable phase under irradiation at high temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Test Plan for the 1600°C Heating Test of AGR-3/4 Compact 3-2

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from the Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 is in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015). This test plan describes the heating test of AGR-3/4 Compact 3-2. The heating test will involve heating the compact in the Fuel Accident Condition Simulator (FACS) furnace to a peak temperature of 1600°C while collecting fission products released from the compact for subsequent analysis. One of several features of AGR-3/4 that set it apart from AGR-1 and AGR-2 is the incorporation of 20 designed-to-fail (DTF) particles in each compact. The fuel kernels in the DTF particles were coated only with a thin (20 µm thick) pyrocarbon layer having high anisotropy such that the coating should most-likely fail during the irradiation, resulting in up to 20 exposed kernels per compact. One of the main goals of this test is to assess the retention of fission products in the exposed kernels and the surrounding graphitic matrix material comprising the compact. Future heating tests will expose AGR-3/4 compacts at different temperatures in the FACS furnace (Demkowicz 2017). Additionally, the ability may be developed to heat re-irradiated AGR-3/4 compacts to allow a determination of the retention of short-lived fission products I-131 and Xe-133.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Test Plan for the 1400°C Heating Test of AGR-3/4 Compact 8-2

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 is in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015). This test plan describes the heating test of AGR-3/4 Compact 8-2. The test will involve heating the compact in the Fuel Accident Condition Simulator (FACS) furnace to a peak temperature of 1400°C while collecting fission products released from the compact for subsequent analysis. One of several features of AGR-3/4 that set it apart from AGR-1 and AGR-2 is the incorporation of 20 designed-to-fail (DTF) particles in each compact. The fuel kernels in the DTF particles were coated only with a thin (20 µm thick) pyrocarbon layer having high anisotropy such that the coating would most-likely fail during the irradiation, resulting in up to 20 exposed kernels per compact. One of the main goals of this test is to assess the retention of fission products in the exposed kernels and the surrounding graphitic matrix material comprising the compact. Future heating tests will expose a number of AGR-3/4 compacts at different temperatures in the FACS furnace (Demkowicz 2017). Additionally, the ability may be developed to heat re-irradiated AGR-3/4 compacts to allow a determination of the retention of short-lived fission products I-131 and Xe-133.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Test Plan for the 1200°C Heating Test of AGR-3/4 Compact 10-2

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 is in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015). This test plan describes the heating test of AGR-3/4 Compact 10-2. The test will involve heating the compact in the Fuel Accident Condition Simulator (FACS) furnace to a peak temperature of 1200°C while collecting fission products released from the compact for subsequent analysis. One of several features of AGR-3/4 that set it apart from AGR-1 and AGR-2 is the incorporation of 20 designed-to-fail (DTF) particles in each compact. The fuel kernels in the DTF particles were coated only with a thin (20 µm thick) pyrocarbon layer having high anisotropy such that the coating would most-likely fail during the irradiation, resulting in up to 20 exposed kernels per compact. One of the main goals of this test is to assess the retention of fission products in the exposed kernels and the surrounding graphitic matrix material comprising the compact. Future heating tests will expose a number of AGR-3/4 compacts at different temperatures in the FACS furnace (Demkowicz 2017). Additionally, the ability may be developed to heat re-irradiated AGR-3/4 compacts to allow a determination of the retention of short-lived fission products I-131 and Xe-133.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

AGR-5/6/7 Irradiation Monitoring Web Site Design

The Nuclear Data Management and Analysis System (NDMAS) manages irradiation experiment data collected under the Advanced Gas Reactor (AGR) Fuel Development and Qualification program within the Advanced Reactor Technologies Technology Development Office. The AGR-5/6/7 irradiation experiment combines the fifth, sixth, and seventh planned irradiation campaigns of the U.S. Department of Energy AGR Fuel Development and Qualification Program. These combined irradiation tests will consist of fuel qualification tests (AGR-5/6) and a margin test (AGR-7) for the industrially produced uranium oxycarbide (UCO) tristructural isotropic (TRISO) coated particle fuel developed by the AGR program. The objective of the AGR-5/6/7 irradiation experiment is to establish the performance of UCO TRISO fuel fabricated using engineering-scale equipment in a production environment under and beyond normal operating conditions (PNL-5245). Experimental irradiation measurements for this experiment (i.e., thermocouple temperature, gas flow rate, fission product release rate, gross gamma count, and Advanced Test Reactor (ATR) operational parameters) are captured automatically to the NDMAS database. Data derived from numerical simulation or other mathematical analyses (i.e., fast fluence, burnup, temperature, and fission product release to birth ratio) are captured to the database when provided by the researchers. The NDMAS database is a relational database accessed via structured query language and managed with Microsoft SQL Server. While select users can directly access the database, that is not the most convenient way to interact with the data. The NDMAS team constructs numerous tables that combine related datasets, and delivers those data to researchers via graphs and spreadsheets on a Microsoft SharePoint-based website. For the AGR-5/6/7 experimental data, a new subsite has been developed, along with a set of web pages. These pages will be used by members of fuel irradiation team for monitoring irradiation conditions according to specification (SPC-1749), by the physics modelers to access necessary model inputs, and by the nuclear engineering community to view and access data such as fuel temperatures, fission product release from capsules, and ATR operational parameters. The deliverable represented in this document is a SharePoint website containing tables of the AGR-5/6/7 measured and calculated data, graphs displaying the data, and links to source documents and references.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

THERMAL ANNEALING OF IRRADIATED PREHYDRIDED ZIRCALOY-4 MATERIALS

Hydrogen-assisted irradiation growth is an important issue for reliability and performance of nuclear materials in nuclear fuel cladding and fuel assembly components. The mechanisms of irradiation growth are not entirely clear and the role of hydrogen and hydrides are less so. There is evidence that breakaway irradiation growth is strongly influenced by the formation and dynamics of dislocation loops that vary with alloy composition, heat treatment and hydrogen content. Thermal annealing of eight Zircaloy-4 specimens irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory as part of the joint EPRI Nuclear Fuels Industry Research (NFIR) and Fuel Reliability Program (FRP) AsTeR irradiation growth study were performed by hot-stage transmission electron microscopy (TEM), differential scanning calorimetry (DSC), and dilatometry via NSUF RTE 17-1093. Two fluence levels (6.7 and 12.3 dpa) and three hydrogen contents (~10, 40, 120 ppm) were investigated. TEM with a hot stage was used for the quantification of "a" and "c" dislocation loop population densities and size distributions at a range of temperatures from 25 to 800oC. DSC and dilatometry were performed by dynamically heating and cooling from 30 to 750ºC at rates of 5, 10, and 20 oC/min to determine the behavior of dislocation loop annealing. Dynamic temperature DSC tests were able to detect the temperature ranges and enthalpies of "a" and "c" loop annihilation, hydride dissolution and precipitation, secondary particle precipitation, and a?a + ß phase transformations upon heating and cooling. Dilatometry tests showed that all of the test showed initial linear expansion followed by a decrease in expansion rate from about 400 to 600 ºC associated with annihilation of "a" and "c" loops and showed evidence for the thermal stabilizing effect of hydrogen on dislocation loops. The results are discussed in the context of the larger AsTeR program that will include higher exposure (~20 dpa) and different alloys when they become available in 2021.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Coordination of Plutonium Separations

NASA and the Department of Energy (DOE) are currently working together to reestablish a 238Pu supply capability using DOE's existing facilities and reactors. The effort is being led by Oak Ridge National Laboratory (ORNL). Target rods, containing neptunium oxide (NpO2), are fabricated at ORNL and irradiated in the ORNL High Flux Isotope Reactor (HFIR) currently and soon will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Irradiated targets are processed in hot cells at the ORNL Radiochemical Engineering Center to recover the plutonium (Pu) product and unconverted neptunium (Np) for recycle. The 238PuO2 product is shipped to Los Alamos National Laboratory (LANL) for fabrication of heat source pellets. Key activities, such as the transport of the neptunium to ORNL, irradiation of neptunium, and chemical processing to recover the newly generated 238Pu, have been demonstrated with the initial amounts of ~100 g heat-source plutonium oxide (HS PuO2) produced. Product samples have been shipped to LANL for evaluation, including chemical impurity analysis. This paper will examine the status of producing new 238Pu from irradiated targets.

Wham, Robert↗