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At least 433 records · Page 24

Corrosion of Aluminum Clad Spent Nuclear Fuel in the 70 Ton Cask during Transfer from L Area to H-Canyon

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Corrosion of Aluminum Clad Spent Nuclear Fuel in the 70 Ton Cask during Transfer from L Area to H-Canyon (Rev.1)

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Joint Workplan on Filler Investigations for DPCs.

This workplan addresses filler attributes (i.e., possible requirements), assumptions needed for analysis, selection of filler materials, testing needs, and a long-range perspective on R&D activities leading to filler demonstration and a safety basis for implementation.

36 MATERIALS SCIENCE↗

Qualifying the Instrument Thimble 11 Test Position in the Advanced Test Reactor Critical

The Advanced Test Reactor (ATR) is a versatile nuclear research reactor located at the Idaho National Laboratory (INL) in Eastern Idaho, United States of America. It is one of the most powerful and flexible research reactors in the world and is primarily used for materials testing, isotope production and basic nuclear science research. The ATR is a light-water-cooled, beryllium-moderated reactor with a nominal thermal power capacity of 250 MW utilizing specially designed fuel, arranged in a serpentine pattern to create 9 flux traps. The special design of ATR allows for the neutron flux within each of the flux traps providing flexibility for experiments, programs and allowing experiments with different dose requirements to be irradiated simultaneously. The Advance Test Reactor Critical (ATRC) is a full-scale replica of the ATR core but is housed in a pool instead of a pressure vessel. ATRC is usually operated at less than 600 watts [1]. There are 12 dry instrument thimbles located around the outside of the core tank where a variety of measurement equipment is housed to facilitate reactor operations. Currently Instrument Thimble 11 (IT-11) is not utilized in ATR or ATRC. A range of measurements is ongoing to qualify IT-11 as a test platform for nuclear instrumentation research, development and further experimentation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Initial Findings in Qualifying the Instrument Thimble 11 Position in the Advanced Test Reactor Critical

The Advanced Test Reactor (ATR) is a versatile nuclear research reactor located at the Idaho National Laboratory (INL) in Eastern Idaho, United States of America. It is one of the most powerful and flexible research reactors in the world and is primarily used for materials testing, isotope production and basic nuclear science research. The ATR is a light-water-cooled, beryllium-moderated reactor with a nominal thermal power capacity of 250 MW utilizing specially designed fuel, arranged in a serpentine pattern to create 9 flux traps. The special design of ATR allows for the neutron flux within each of the flux traps providing flexibility for experiments, programs and allowing experiments with different dose requirements to be irradiated simultaneously. The Advance Test Reactor Critical (ATRC) is a full-scale replica of the ATR core but is housed in a pool instead of a pressure vessel. ATRC is usually operated at less than 600 watts [1]. There are 12 dry instrument thimbles located around the outside of the core tank where a variety of measurement equipment is housed to facilitate reactor operations. Currently Instrument Thimble 11 (IT-11) is not utilized in ATR or ATRC. A range of measurements is ongoing to qualify IT-11 as a test platform for nuclear instrumentation research, development and further experimentation.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady-state heat fluxes of 5–20 MW/m 2 and ion fluxes up to 10 24 m -2 s -1 . Since PFCs will have to withstand neutron irradiation displacement damage up to 50 dpa, the target station design must accommodate radioactive specimens (materials to be irradiated in HFIR or at SNS) to enable investigations of the impact of neutron damage on materials. Therefore, the system will have to be able to install and extract irradiated specimens using equipment and methods to avoid sample modification, control contamination, and minimize worker dose. Included in the design considerations will be an assessment of all the steps between neutron irradiation and post-exposure materials examination/characterization, as well as an evaluation of the facility hazard categorization. In particular, the factors associated with the acquisition of radioactive specimens and their preparation, transportation, experimental configuration at the plasma-specimen interface, post-plasma-exposure sample handling, and specimen preparation will be evaluated. Neutronics calculations to determine the dose rates of the samples were carried out for a large number of potential plasma-facing materials.

42 ENGINEERING↗

Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices that contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify nuclear heating. The last one consists in the development of accurate measurement and analysis methods. The paper will be dedicated to a complete review of the experimental and numerical works performed since 2009 thanks to two parts. The first part will detail a new thermal approach implemented to improve nuclear heating measurements by radiometric calorimeters. New experimental tools (calorimeter prototypes and set-ups such BETHY Bench) developed to perform preliminary out-of-pile studies under suitable conditions will be presented (temperature and velocity of the external cooling fluid, heat source localization and intensity inside the calorimetric cells). Then the response of two kinds of sensors, their calibrations curves and their thermal behaviors will be compared for various parameters. Finally validated numerical thermal and Monte Carlo works will be discussed to propose new improvements. The second parts of the paper will focus on works realized in order to design, develop and test the first prototype of the multi-sensor device called CARMEN [7-9]. The two mock-ups dedicated respectively to neutron measurements and photon measurements will be detailed. The results obtained during two irradiation campaigns inside the periphery of OSIRIS reactor will be shown. The new analysis method will be discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Accelerated Materials Deployment in Advanced Nuclear Power Plants

The purpose of this report is to begin the development of a maximally efficient process for licensing and deploying new materials in Advanced Non-Light-Water Reactors (ANLWRs). Some new materials that are to be used in some new plants are seen as possibly introducing risks, because our understanding of those new materials’ behavior in the conditions generated by some novel plant designs is less complete than our understanding of the behavior of materials with long use histories in existing designs. In these cases, an approved code/standard or a code case to support the use of these materials in the novel design’s safety case may not exist for the regulator to utilize as part of the licensing determination. This circumstance creates the potential for an extremely long licensing process for new designs using new materials. The present strategy is to show how to manage these risks proactively, in such a way as to permit licensing decisions to be made in a timely manner, based on this risk management process. The present report outlines the gaps in the current codes to support deployment and use of novel materials and begins the development of the necessary risk management framework that is focused on the subject materials issues; it is based on risk-informed in-service surveillance practices, carried out in such a way as to compensate for current limitations in our state of knowledge. This development will enable licensing and deployment of the subject materials, conditional on the proactive surveillance process to be established. While this report is occasioned by limitations in our knowledge of certain materials issues that may arise in advanced designs, in-service surveillance is always done in order to compensate for a lack of knowledge: if we knew that components were not already failed and not trending toward failure, we would not perform surveillance, even in current-generation plants (except that prescriptive requirements would force us to do so). What is different about the surveillance program discussed here is that the issues are newer and the relevant experience base is less complete, so the surveillance presently contemplated may need to measure new things and/or measure them more often than has been traditional for surveillance coupons. The present report is devoted to the risk management framework and applies American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Division [1] to establish the structure of a protocol for carrying out the necessary surveillance. These documents are generic: they do not tell us how often to surveille, or what to surveille, or what to measure, but rather how to determine those things, given certain technical inputs. The Regulatory Development R&D Program [2] is currently developing the companion supporting technical basis for the materials surveillance technology that, when completed and validated, can be used by owner/operator and NRC to implement a materials degradation management program for ANLWRs. This report also outlines salient points of discussion, positive potential outcomes, and potential concerns from industry and the USNRC. These aspects of the report intend to inform future work to develop a proposed technical process for adoption by the industry and endorsement by the USNRC to allow developers to propose a risk informed and conservative approach for the use of materials where operating experience/data and codes and standards may not exist for use of a novel material in an operating reactor environment. Additionally, such a technology could be leveraged to potentially reduce part of the upfront materials data requirements from ongoing long-term materials testing so that early action on license application could be undertaken by NRC, in parallel with the continuation of long-term data collection. This could accelerate the schedule for a first-of-a-kind ANLWR deployment or a nth-of-a-kind new materials insertion for established ANLWR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Data-driven design of metal–organic frameworks for wet flue gas CO 2 capture

Limiting the increase of CO 2 in the atmosphere is one of the largest challenges of our generation. Because carbon capture and storage is one of the few viable technologies that can mitigate current CO 2 emissions, much effort is focused on developing solid adsorbents that can efficiently capture CO 2 from flue gases emitted from anthropogenic sources. One class of materials that has attracted considerable interest in this context is metal-organic frameworks (MOFs), in which the careful combination of organic ligands with metal-ion nodes can, in principle, give rise to innumerable structurally and chemically distinct nanoporous MOFs. However, many MOFs that are optimized for the separation of CO 2 from nitrogen do not perform well when using realistic flue gas that contains water, because water competes with CO 2 for the same adsorption sites and thereby causes the materials to lose their selectivity. Although flue gases can be dried, this renders the capture process prohibitively expensive. Here we show that data mining of a computational screening library of over 300,000 MOFs can identify different classes of strong CO 2 -binding sites-which we term 'adsorbaphores'-that endow MOFs with CO 2 /N 2 selectivity that persists in wet flue gases. We subsequently synthesized two water-stable MOFs containing the most hydrophobic adsorbaphore, and found that their carbon-capture performance is not affected by water and outperforms that of some commercial materials. Testing the performance of these MOFs in an industrial setting and consideration of the full capture process-including the targeted CO 2 sink, such as geological storage or serving as a carbon source for the chemical industry-will be necessary to identify the optimal separation material.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Propagation of a short-pulse laser-driven electron beam in matter

We studied the transport of an intense electron beam produced by high intensity laser pulses through metals and insulators. Targets were irradiated at two different intensities, 10{sup 17} W/cm{sup 2} and 10{sup 19} W/cm{sup 2}, at the laser facility Xtreme Light XL-III in Beijing, a Ti:Sa laser source emitting 40 fs pulses at 800 nm. The main diagnostic was Cu-K{sub {alpha}} fluorescence imaging. Images of K{sub {alpha}} spots have been collected for those two laser intensities, for different target thickness, and for different materials. Experimental results are analyzed taking into account both collisional and collective effects as well as refluxing at the edge of the target. The target temperature is evaluated to be T{sub c} {approx} 6 eV for intensity I = 10{sup 17} W/cm{sup 2} (for all the tested materials: plastic, aluminium, and copper), and T{sub c} {approx} 60 eV in aluminium and 120 eV in titanium for intensity I = 10{sup 19} W/cm{sup 2}.

36 MATERIALS SCIENCE↗

Understanding and Predicting Effect of Sodium Exposure on Microstructure of Grade 91 Steel

This report provides an update on the understanding of the effect of sodium exposures on microstructure and tensile properties of Grade 91 (G91) steel in support of the design and operation of G91 components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602018), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Application of Proliferation Resistance Optimization (PRO-X) Methodology to a Generic Research Reactor

The Proliferation Resistance Optimization Program (PRO-X) has been established by the NNSA to provide a framework for evaluating and integrating proliferation resistance into nuclear reactor system (core, fuel, and auxiliary facilities) designs that also maintain the safety and peaceful use missions of those systems. The research reactor (PRO-RR) area is the component of PRO-X that supports the program objectives by applying state-of-the-art analysis methods to research reactor systems. This report details the results of using a defined set of analytical tools to evaluate the neutronics, thermal hydraulics and proliferation risk characteristics of a set of parametric cores based on a generic 10 MW materials test reactor that uses a plate-type low enriched uranium (LEU) fuel. The analysis shows that by suitable adjustment of the core size, reflector configuration and power level, mission performance and safety margins can be maintained or improved while reducing the potential for production of special nuclear material.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

State of the art, gaps, and prospects in fusion materials theory and modelling

Advancing the theory and simulation of materials for fusion applications remains a key component of global roadmaps aimed at delivering much-needed fusion power. Especially as the drive for commercial application increases, prototypes must be designed against radiation damage before the relevant experimental data can be collected and cost reductions that are possible by testing materials in silico become even more important. Here, we summarise the state of the art as it emerged during the 7 th Fusion Materials Theory & Modelling Workshop that took place in 2024, with the aim to highlight present gaps and future directions for the fusion materials modelling community. Of particular interest were the effects of transmutations, chemical complexity with the development of novel alloys and interatomic potentials, advancements in modelling high-dose microstructures, comparison with experimental data and multiscale models for structural assessment relying on high-performance computing and virtual reality.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Fuels Characterization for National Research Council Canada 2-m Pool Fire Test Series

This report provides a detailed analysis of the physical and chemical properties of three liquid hydrocarbon fuels: heptane, Bakken crude, and a diluted bitumen, that were subsequently tested in a series of 2-m pool fire experiments at Sandia National Laboratories for the National Research Council Canada. Properties such as relative density, vapor pressure (VPCR x ), composition, and heat of combustion were evaluated. The heptane analysis, with relative density = 0.69 (at 15°C), confirmed that the material tested was consistent with high-purity (>99%) n-heptane. The Bakken crude, with a relative density = 0.81 (at 15°C), exhibited a vapor pressure by VPCR 0.2 (37.8°C) in the range 120-157 kPa. The dilbit, with a relative density = 0.92 (at 15°C) exhibited a vapor pressure by VPCR 0.2 (37.8°C) in the range 85-98 kPa. Solids remaining in the test pans after the pool fires were also collected and weighed. No detectable solids were left after the heptane burns. In contrast, the crude oils left some brittle, black solid residue. On average, dilbit pool fires left about 40 more residue by mass than Bakken pool fires for equivalent mass of fuel feed.

01 COAL, LIGNITE, AND PEAT↗

Development and Execution of a Large-scale DDT Tube Test for IHE Material Qualification

Insensitive High Explosive (IHE) Materials are defined in Chapter IX of the DOE Explosive Safety Standard (DOE-STD-1212-2012) as being materials that are massdetonable explosives that are so insensitive that the probability of accidental initiation or transition from burning to detonation is negligible1. There are currently a number of tests included in the standard that are required to qualify a material as IHE, however, none of the tests directly evaluate for the transition from burning to detonation (aka deflagration-to-detonation transition, DDT). Currently, there is a DOE complex-wide effort to revisit the IHE definition in DOE-STD-1212-2012 and change the qualification requirements. The proposal lays out a new approach, requiring fewer, but more appropriate tests, for IHE Material qualification. One of these new tests is the Deflagration-to-Detonation Test. According to the redefinition proposal, the purpose of the new deflagration-todetonation test is “to demonstrate that an IHE material will not undergo deflagration-to-detonation under stockpile relevant conditions of scale, confinement, and material condition. Inherent in this test design is the assumption that ignition does occur, with onset of deflagration. The test design will incorporate large margins and replicates to account for the stochastic nature of DDT events.” In short, the philosophy behind this approach is that if a material fails to undergo DDT in a significant over-test, then it is extremely unlikely to do so in realistic conditions. This effort will be valuable for the B61 LEP to satisfy their need qualify the new production lots of PBX 9502. The work described in this report is intended as a preliminary investigation to support the proposed design of an overly conservative, easily fielded DDT test for updated IHE Material Qualification standard. Specifically, we evaluated the aspects of confinement, geometry, material morphology and temperature. We also developed and tested a thermally robust igniter system.

36 MATERIALS SCIENCE↗

SU-E-J-210: Characterizing Tissue Equivalent Materials for the Development of a Dual MRI-CT Heterogeneous Anthropomorphic Phantom Designed Specifically for MRI Guided Radiotherapy Systems

Purpose: MRI guided radiotherapy (MRIgRT) is an emerging technology which will eventually require a proficient quality auditing system. Due to different principles in which MR and CT acquire images, there is a need for a multi-imaging-modality, end-to-end QA phantom for MRIgRT. The purpose of this study is to identify lung, soft tissue, and tumor equivalent substitutes that share similar human-like CT and MR properties (i.e. Hounsfield units and relaxation times). Methods: Materials of interested such as common CT QA phantom materials, and other proprietary gels/silicones from Polytek, SmoothOn, and CompositeOne were first scanned on a GE 1.5T Signa HDxT MR. Materials that could be seen on both T1-weighted and T2-weighted images were then scanned on a GE Lightspeed RT16 CT simulator and a GE Discovery 750HD CT scanner and their HU values were then measured. The materials with matching HU values of lung (−500 to −700HU), muscle (+40HU) and soft tissue (+100 to +300HU) were further scanned on GE 1.5T Signa HDx to measure their T1 and T2 relaxation times from varying parameters of TI and TE. Results: Materials that could be visualized on T1-weighted and T2-weighted images from a 1.5T MR unit and had an appropriate average CT number, −650, −685, 46,169, and 168 HUs were: compressed cork saturated with water, Polytek Platsil™ Gel-00 combined with mini styrofoam balls, radiotherapy bolus material, SmoothOn Dragon-Skin™ and SmoothOn Ecoflex™, respectively. Conclusion: Post processing analysis is currently being performed to accurately map T1 and T2 values for each material tested. From previous MR visualization and CT examinations it is expected that Dragon-Skin™, Ecoflex™ and bolus will have values consistent with tissue and tumor substitutes. We also expect compressed cork statured with water, and Polytek™-styrofoam combination to have approximate T1 and T2 values suitable for lung-equivalent materials.

60 APPLIED LIFE SCIENCES↗

Micro- and macroscale coefficients of friction of cementitious materials

Millions of metric tons of cementitious materials are produced, transported and used in construction each year. The ease or difficulty of handling cementitious materials is greatly influenced by the material friction properties. In the present study, the coefficients of friction of cementitious materials were measured at the microscale and macroscale. The materials tested were commercially-available Portland cement, Class C fly ash, and ground granulated blast furnace slag. At the microscale, the coefficient of friction was determined from the interaction forces between cementitious particles using an Atomic Force Microscope. At the macroscale, the coefficient of friction was determined from stresses on bulk cementitious materials under direct shear. The study indicated that the microscale coefficient of friction ranged from 0.020 to 0.059, and the macroscale coefficient of friction ranged from 0.56 to 0.75. The fly ash studied had the highest microscale coefficient of friction and the lowest macroscale coefficient of friction. -- Highlights: •Microscale (interparticle) coefficient of friction (COF) was determined with AFM. •Macroscale (bulk) COF was measured under direct shear. •Fly ash had the highest microscale COF and the lowest macroscale COF. •Portland cement against GGBFS had the lowest microscale COF. •Portland cement against Portland cement had the highest macroscale COF.

36 MATERIALS SCIENCE↗

Challenge of Fielding Experiments at the Annular Core Research Reactor

Conference is the annual Test, Research &Training Reactors (TRTR) meeting. Typically attended by dozens of research reactor operators, educators, trainers, maintenance & service contractors Also, members of the Nuclear Regulatory Commission (NRC) attend. Presentation highlights ACRRs capability and the process used to submit and approve experimental activities. This presentation is adapted from my previous presentation to the Material Test and Research conference, SAND2025-1775526.

Martin, Lonnie E. [Sandia National Laboratories (S↗