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At least 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

A Continuum Dislocation Dynamics Crystal Plasticity Approach to Irradiated Body-Centered Cubic α-Iron

Radiation-induced embrittlement of reactor pressure vessel (RPV) steels can potentially limit the operating life of nuclear power plants. Over extended exposure to radiation doses, these body-centered cubic (BCC) irons demonstrate irradiation damage. Here, we present a continuum dislocation density (CDD) crystal plasticity model to capture the interaction among dislocations and self-interstitial atom (SIA) loops in α-iron. We demonstrate the importance of modeling cross slip using a combined stochastic Monte Carlo approach and the role of slip system strength anisotropy in capturing stochastic cross slip interactions. Through these captured interactions, the CDD crystal plasticity model can capture both the stress response and the physical evolution of dislocations on different slip system planes. Single-crystal verification experiments are used to calibrate the CDD crystal plasticity model, and a set of simplified polycrystalline simulations demonstrates the model’s ability to capture the stress response from tensile experiments on α-iron.

36 MATERIALS SCIENCE↗
DOE OSTI2021

IN-SERVICE INSPECTION OF EXTENDED DRY STORAGE OF SPENT NUCLEAR FUEL, PART I: CRAWLER TECHNOLOGY DEVELOPMENT

This paper describes development and demonstration of remote crawling systems to support periodic examinations of interim dry storage system (DSS) canisters for spent nuclear fuel in the USA. Specifically, this work relates to robotic crawler developments for “canister” based DSS systems which form the majority population of DSSs, in the USA, for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond their initial licensed or certified terms. Challenges associated with performing the periodic examinations are associated with physical access to the canister surface which is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics. Motivations for performing periodic examinations and developing robotic crawlers for performing those examinations remotely will be presented and several activities to demonstrate robotic crawlers for different DSS systems are summarized.

nondestructive examination (NDE), Dry Cask Storage↗
DOE OSTI2021

INSERVICE INSPECTION OF EXTENDED DRY STORAGE OF SPENT NUCLEAR FUEL, PART II: NDE/SENSOR TECHNOLOGY DEVELOPMENT AND CODIFICATION

This paper describes development and demonstration of nondestructive examination (NDE) technologies to support periodic examinations of interim dry cask storage system (DCSS) canisters for spent nuclear fuel in the USA to verify continued safe operation and that the canister confinement is intact and performing its intended safety function. Specifically, this work relates to NDE technology development for “canister” based DCSS systems which form the majority population of DCSSs, in the USA, for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond the initial license or certified term. Physical access to the canister surface is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics related to inspections. Several activities to demonstrate NDE technologies for the inspections of different DCSS systems are summarized.

nondestructive examination (NDE), Dry Cask Storage↗
DOE OSTI2025

Black-box optimization of CT acquisition and reconstruction parameters: a reinforcement learning approach

Protocol optimization is critical in Computed Tomography (CT) for achieving desired diagnostic image quality while minimizing radiation dose. Due to the inter-effect of influencing CT parameters, traditional optimization methods rely on the testing of exhaustive combinations of these parameters. This poses a notable limitation due to the impracticality of exhaustive parameter testing. This study introduces a novel methodology leveraging Virtual Imaging Trials (VITs) and reinforcement learning to more efficiently optimize CT protocols. Computational phantoms with liver lesions were imaged using a validated CT simulator and reconstructed with a novel CT reconstruction Toolkit. The optimization parameter space included tube voltage, tube current, reconstruction kernel, slice thickness, and pixel size. The optimization process was done using a Proximal Policy Optimization (PPO) agent which was trained to maximize the Detectability Index (d’) of the liver lesion for each reconstructed image. Results showed that our reinforcement learning approach found the absolute maximum d’ across the test cases while requiring 79.7% fewer steps compared to an exhaustive search, demonstrating both accuracy and computational efficiency, offering a efficient and robust framework for CT protocol optimization. The flexibility of the proposed technique allows for use of varying image quality metrics as the objective metric to maximize for. Our findings highlight the advantages of combining VIT and reinforcement learning for CT protocol management.

Fenwick, David [Duke University Medical Center]↗
DOE OSTI2020

Development of Cable Aging Acceptance Criteria for Nuclear Facilities

The aging of nuclear power plant electrical cables has been the subject of substantial research and development (R&D) projects in the nuclear industry, national and international laboratories, universities, and vendor organizations for several years. More specifically, these R&D efforts were conducted to develop equipment and techniques for assessing age-related degradation of in-service cables in nuclear facilities including power plants, research reactors, waste facilities and fuel fabrication plants. In recent years, cable R&D efforts in the nuclear arena have also included work to develop new cables for the next generation of reactors. Over time, exposure to harsh environmental conditions such as elevated temperatures, radiation, and humidity in nuclear installations can result in age-related degradation and failure of cables. In the current fleet of nuclear reactors, there are thousands of miles of cabling installed in each plant and many of these cables are exposed to these harsh environmental conditions. For these cables, the jacket and insulation polymers harden and become brittle over time, making them more susceptible to crack formation and growth, moisture intrusion, and other mechanisms that can lead to cable failure. Moreover, the existing U.S. fleet of 98 nuclear reactors has an average operating age of 38 years. Many of these nuclear power plants have applied for and almost all have been granted regulatory approval for license renewals to operate for 60 years, 20 years beyond their original 40-year life. Further, subsequent license renewals (SLRs) are underway with a few commercial nuclear power sites already approved to operate up to 80 years. As these reactors pursue operating life extensions, utilities must find a way to address issues associated with age-related degradation of cables. Designers of small modular reactors are also in need of technologies to evaluate the performance of cables that will be installed in harsher environments (e.g. higher temperatures, radiation doses, dose rates, etc.) than those present in current-generation reactors. Today, a variety of cable condition monitoring (CM) techniques have been developed and successfully used in nuclear facilities. These techniques are used to identify age-related degradation and assess the condition of cables to determine if their performance characteristics have changed with age. However, objective criteria must be developed for these CM tests to help quantify cable condition and thereby develop repair and replacement schedules. With operating life extending to 80 years and more SLRs on the horizon, both nuclear facilities and regulators need an objective means to determine the aged condition of cables.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Stochastic parametric skeletal dosimetry model for humans: Anatomical-morphological basis and parameter evaluation

Radiation exposure of the hematopoietic system that results in a radiation dose to bone marrow of more than 100 mGy leads to an increase in the risk of leukemia in humans. Excess relative risk of leukemia was observed in cohorts whose members lived in the territories of the Southern Urals that were radioactively contaminated in the 1950s. As part of the dosimetric support of epidemiological studies of these cohorts, an original methodology for stochastic bone dosimetric modeling was developed, termed the Stochastic Parametric Skeletal Dosimetry (SPSD) model. The purpose of this work was to present the anatomical and morphological bases of the SPSD model, which includes an assessment of the parameters of the microstructure of the trabecular bone in the hematopoietic areas of the human skeleton, as well as a description of the macrostructural division (segmentation) of hematopoietic areas into simple bone segments. As a result, an anatomical-morphological basis of the SPSD model was created based on published data. Data collection work included the analysis of original articles, atlases, manuals, monographs and the formation of primary data files. Data on the duration of hematopoiesis in various parts of the skeleton; and data on age-related changes in the microstructure and linear dimensions of human bones and their segments were analyzed. The paper describes the full set of parameters to be used for the dosimetric model for newborns, children aged 1, 5 and 10 years, as well as for adolescents aged 15 years and adults; for the latter, sex differences in bone size were considered. In total, the SPSD model includes 289 unique basic (bone) phantom segments, each of which is described by 7 or more parameters describing the microstructure, thickness of the cortical layer and linear dimensions. Population variability was estimated for each parameter. The approach to SPSD modeling, i.e., the use of simple geometric shapes, was successfully verified using independent datasets on bone masses and volumes.

Science & Technology - Other Topics↗
DOE OSTI2024

Progress in the design of the future circular collider FCC-ee interaction region

In this paper we discuss the latest developments for the FCC-ee interaction region layout, which represents one of the key ingredients to establish the feasibility of the FCC-ee. The collider has to achieve extremely high luminosities over a wide range of center-of-mass energies with two or four interaction points. The complex final focus hosted in the detector region has to be carefully designed, and the impact of beam losses and of any type of synchrotron radiation generated in the interaction region, including beamstrahlung, have to be evaluated in detail with simulations. We give an overview of the progress of the whole machine-detector-interface-related studies, among which are the updated mechanical model of the interaction region, the plans for a novel R&D activity of a IR mockup which is just starting, the collimation scheme and evaluation of beam induced backgrounds in the detectors, evaluation of radiation dose in the experimental area, and MDI integration with the detector.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

Characterization of CST Post-Processing AP-105 Hanford Tank Waste

The primary goal of the Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions, LLC (WRPS), is to remove entrained solids and 137 Cs from the Hanford tank waste supernate to expedite production of low-activity waste. Ion exchange (IX) testing of 10.9 L of waste from Hanford tank 241-AP-105 (AP-105), performed by Pacific Northwest National Laboratory, used a lead-lag-polish column format, with a bed volume of 10 mL per column, to decontaminate tank waste supernate using crystalline silicotitanate (CST) as the IX media. The AP-105 Cs IX processing test, discussed elsewhere, resulted in a shorter transition zone (i.e., steeper load curve) than those defined by wastes from tanks 241-AP-107 and 241-AW-102.1 The shorter transition zone was indicative of a matrix effect retarding Cs capacity. Therefore, aliquots of spent CST from the lead, lag, and polish columns were subjected to a digestion protocol to quantify analytes retained by the CST and extrapolate the impact on Cs capacity. The spent CST was digested using a combination of 5 M HNO 3 and H 2 O 2 with vigorous heating and stirring. Due to the radiation dose accompanying the 137 Cs on the CST columns, a secondary Cs separation by ammonium molybdophosphate embedded in polyacrylonitrile (AMP-PAN) was performed to separate the 137 Cs from the CST so the samples could be contact-handled for analysis outside of a shielded facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Development and Advanced Testing of DPC Cement Fillers

The fillers research and development (R&D) program, mostly experimental, is part of a broader R&D program that includes new process modeling and performance assessment of criticality effects and the overall importance of criticality to repository performance (consequence screening). A literature research and consultation effort with experts by Hardin and Brady (2018) identified several potentially effective and workable filler materials including cements (primarily phosphate based), moltenmetal alloys, and low-temperature glasses. Filler attributes were defined, and the preliminary lists were compared qualitatively. Further comparative analysis will be done (e.g., cost estimates) after experimental screening has narrowed the list of alternatives. The research presented here is focused Sandia’s efforts for the development of phosphate-based cement fillers. Molten metal filler research is an ongoing activity at Oak Ridge National Laboratories and is not discussed herein. After the completion of the FY20 research effort the following cement filler compositions were selected for further experimental development work and advanced testing in FY21: 1. Aluminum phosphate cements (APCs); more specifically aluminum oxide / aluminum phosphate (Al 2 O 3 / AlPO 4 ) cements in which Al 2 O 3 serves as the filler material bound by an AlPO 4 binder formed by the reaction of Al 2 O 3 with various phosphate sources; 2. Wollastonite phosphate cements (WPCs), specifically wollastonite and aluminum or calcium aluminum phosphates in which CaSiO 3 serves as the filler material bound by a calcium phosphate that serves as the binder; and 3. Calcium aluminate phosphate cements (CAPCs) specifically grossite (CaAl 4 O 7 ) and hibonite (CaAl 11 O 18 ) fillers bound by an aluminum phosphate that serves as the binder. This effort focused on the optimization and subsequent processing of these three cements to achieve dense and well-consolidated monolithic samples. Upon completion of the FY21 effort the aluminum phosphate cements (APCs) and the calcium aluminate phosphate cements (CAPCs) show the most promise for advanced testing and scale up. We will begin the work in FY22 focused on testing the performance of these two cements in small scale DPCs as well as advanced materials testing to evaluate cement performance under expected radiation doses and representative post-closure geochemical environments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Development of Cable Aging Acceptance Criteria for Nuclear Facilities

The aging of nuclear power plant electrical cables has been the subject of substantial research and development (R&D) projects in the nuclear industry, national and international laboratories, universities, and vendor organizations for several years. More specifically, these R&D efforts were conducted to develop equipment and techniques for assessing age-related degradation of in-service cables in nuclear facilities including power plants, research reactors, waste facilities and fuel fabrication plants. In recent years, cable R&D efforts in the nuclear arena have also included work to develop new cables for the next generation of reactors. Over time, exposure to harsh environmental conditions such as elevated temperatures, radiation, and humidity in nuclear installations can result in age-related degradation and failure of cables. In the current fleet of nuclear reactors, there are thousands of miles of cabling installed in each plant and many of these cables are exposed to these harsh environmental conditions. For these cables, the jacket and insulation polymers harden and become brittle over time, making them more susceptible to crack formation and growth, moisture intrusion, and other mechanisms that can lead to cable failure. Moreover, the existing U.S. fleet of 98 nuclear reactors has an average operating age of 38 years. Many of these nuclear power plants have applied for and almost all have been granted regulatory approval for license renewals to operate for 60 years, 20 years beyond their original 40-year life. Further, subsequent license renewals (SLRs) are underway with a few commercial nuclear power sites already approved to operate up to 80 years. As these reactors pursue operating life extensions, utilities must find a way to address issues associated with age-related degradation of cables. Designers of small modular reactors are also in need of technologies to evaluate the performance of cables that will be installed in harsher environments (e.g. higher temperatures, radiation doses, dose rates, etc.) than those present in current-generation reactors. Today, a variety of cable condition monitoring (CM) techniques have been developed and successfully used in nuclear facilities. These techniques are used to identify age-related degradation and assess the condition of cables to determine if their performance characteristics have changed with age. However, objective criteria must be developed for these CM tests to help quantify cable condition and thereby develop repair and replacement schedules. With operating life extending to 80 years and more SLRs on the horizon, both nuclear facilities and regulators need an objective means to determine the aged condition of cables.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Changes to the morphology and coefficient of thermal expansion in HDPE and UHMWPE following irradiation-based crosslinking

This work describes recent findings of the Los Alamos National Laboratory concerning the alterations to polyethylene’s (PE’s) mechanical and chemical properties after crosslinking via gamma irradiation. Three materials were examined as part of this study, including one HDPE and two UHMWPE specimens. Changes to degree of crystallinity (X C ) were qualitatively determined via measurements of density, and results suggest that X C is not necessarily correlated to crosslink density, depending on polymer manufacturing process and treatment following irradiation. TMA data suggests an anisotropic response in linear thermal expansion, and that the response may be more correlated to X C than crosslink density. Finally, FTIR analysis has led to the conclusion that oxidation occurs as a result of irradiation in open air, the degree of oxidation is proportional to the maximum radiation dose received, and that oxidation is largely a surface effect, penetrating no deeper than 1.5 mm. Due to time constraints, a full dataset could not be generated for changes in creep and thermal properties as a result of irradiation, and TMA analysis is only complete for one material tested in this study. These measurements are ongoing and will provide more context to the hypotheses drawn in this work.

36 MATERIALS SCIENCE↗
DOE OSTI2022

2021 Idaho National Laboratory Site Environmental Report

The INL Site’s operations, as well as the ongoing cleanup, necessarily involve a commitment to environmental stewardship and full compliance with environmental protection laws. As part of this commitment, the INL Site Environmental Report is prepared annually to inform the public, regulators, stakeholders, and other interested parties of the INL Site’s environmental performance during the year. This report is published for the DOE-ID in compliance with DOE O 231.1B, “Environment, Safety and Health Reporting.” Its purpose is to: • Present the INL Site, mission, and programs • Report compliance status with applicable federal, state, and local regulations • Describe the INL Site environmental programs and activities • Summarize results of environmental monitoring • Discuss potential radiation doses to the public residing in the vicinity of the INL Site • Report on ecological monitoring and research conducted by contractors and affiliated agencies and by independent researchers through the Idaho National Environmental Research Park • Describe quality assurance methods used to ensure confidence in monitoring data • Provide supplemental technical data and reports that support the INL Site Environmental Report (https://idahoeser.inl.gov/publications.html).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

Design and Prototyping of a Novel Toroid Magnet System for the MOLLER Experiment at Jefferson Lab

The Thomas Jefferson National Accelerator Facility (JLab) has designed a unique spectrometer system to measure the weak interaction between electrons. The experiment ?Measurement of Lepton-Lepton Electroweak Reaction? (MOLLER) requires leveraging the recent 12 GeV electron beam upgrade and will run in JLab for 3 years. Focusing the signal for the MOLLER experiment requires five water-cooled toroidal magnets, each with unique geometry and with 7-fold symmetry. This system of magnets provides the magnetic field required to separate the incident beam electrons scattered from the target electrons (Møller scattering) and protons (elastic e-p scattering) in a liquid hydrogen target. The conceptual design was developed by the MOLLER collaboration and was given to JLab in the form of amp turns and physical location, with additional physics requirements. This paper presents prototyping of the coils and magnet support system and discusses the lessons learned during the process along with the plans for full magnet testing and installation. The JLab Magnet Group along with the MOLLER collaboration developed the specification document that includes keep out zones to design the set of magnets. JLab contracted the design of the first toroid magnet in the magnet (TM0) to Massachusetts Institute of Technology. The other four toroid magnets (TM1 through TM4) have been designed by JLab and are in the process of fabrication and assembly. Prototype coils of TM1-TM4 have been fabricated by Everson-Tesla Incorporated, PA (USA). The manuscript presents the unique challenges of the design, alignment, high current density, operating range, high radiation dose, and vacuum environment.

Kashy, David↗
DOE OSTI2023

Robotics Plan in Support of DOME Testbed Operations

Various robotic tooling options have been evaluated for the NRIC DOME concept of operations (ConOps). Framatome was contracted to develop a wide-ranging list of commercial off the shelf (COTS) and custom robotic systems to be considered for performing the DOME ConOps functions. Subsequent project tasks from Framatome narrowed down the list and scored the most viable options. The abbreviated list, and associated scoring, has been reviewed and assessed to provide formal recommendations for the robotic ConOps functions of DOME. The overhead telescoping mast & Kraft arm assembly is recommended as the primary system for reactor demobilization and removal. The estimated cost is $\$700$K with a timeline to develop and deliver of about 2.5 years. The mast & Kraft arm would still require an overhead lift system for mobilization and installation. Either the refurbished polar crane or a new gantry crane delivery platform could both serve as the overhead lift and delivery system. The estimated costs for the refurbishment polar crane and new gantry crane systems are, respectively, $\$4.4$M and $\$3$M with about 2.5 years to develop and deliver. A Brokk + Kraft crawler & arm is recommended as a secondary robotic system to support the overhead mast & Kraft arm system. The Brokk + Kraft crawler & arm would cost an estimated $\$726$k and would take about 1.5 years to develop and deliver. The Boston Dynamics SPOT robot is also recommended for the in-service monitoring during experimentation operations. The estimated cost is $\$220$K with a timeline to deliver of 8-14 months. The Elios 3 aerial drone is recommended to perform large area radiation dose rate mapping and visual inspections. The estimated cost is a $\$100$K and will also need about 8-14 months to develop and deliver. A mockup rig is also recommended to be procured an employed to test & verify the robotic capabilities as well as provide needed training for operations personnel. It is estimated that a mockup rig would cost about $\$500$k and would take up to a year to develop and build.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Compact Superconducting RF Accelerators for Water Treatment

Jefferson Lab is actively exploring the environmental applications of compact irradiation facilities utilizing Superconducting Radio Frequency (SRF) accelerator technology ? a well-established technology primarily used in large research machines. Recent advancements, such as the application of thin films of superconducting alloys in radio frequency (RF) resonators and the development of high-capacity cryocoolers, have made it feasible to employ this key technology in compact, standalone irradiation facilities. The main advantage over conventional systems based on normal conducting technology is the significantly enhanced efficiency in beam power generation, enabling exposure to substantially higher radiation doses. In the realm of addressing contaminants in wastewater treatment, particularly concerning persistent substances known as "forever chemicals," this technology presents an attractive solution, given the required considerable effort i.e. dose to break them down. Experiment Numbers:

Vennekate, John↗
DOE OSTI2024

Considerations for Defining G-Values for Aluminum-Clad Spent Nuclear Fuel

Sealed-canister dry storage of aluminum-clad spent nuclear fuel (ASNF) generated by research reactors is an alternative to current storage and disposition pathways as directed by the U.S. Department of Energy. The major challenge faced for this storage approach is radiolytic H 2 generation, including from the aluminum (oxy)hydroxide layers on the surface of ASNF. Experimental and modeling activities have been carried out to characterize the radiolytic yield as part of a DOE-sponsored research program to develop the technical basis for ASNF dry storage. The G-value is a commonly way to report results of radiolysis testing and is defined as the radiolytic yield of a species (e.g. molecular hydrogen) per unit radiation energy deposited into the material system. An independent technical review of the ASNF dry storage technical basis performed by Pacific Northwest National Laboratory raised questions about differences in G-value definitions used for experiments on ASNF surrogates consisting of aluminum samples with adherent (oxy)hydroxides compared to G-values reported in prior literature and how the magnitudes compared between different studies. Material systems resembling ASNF pose complications for measuring/defining G-values to predict the evolution of H 2 in a sealed canister, including i) accounting for radiolytic yields potentially arising from multiple sources, i.e., residual free (vapor), physisorbed, and chemisorbed/chemically bound waters; ii) deciding what portions of the multi-material system to include in the absorbed energy (radiation dose) calculation, considering possible energy exchange between materials as well as measurement limitations, and iii) capturing variations in G-value associated with non-linear yield vs. dose curves and/or dependence on the cover gas. This report summarizes previous literature information on radiolytic H 2 generation and associated G-values from mixed-material systems (generally oxides in contact with water or organic compounds) and from (oxy)hydroxides/hydrates to compare with the definitions and values for ASNF surrogate samples containing adherent aluminum (oxy)hydroxides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

LUCID Thrust 1 - Dataset Identification and Biodata Catalog Creation

The LUCID DOE consortium, part of the Department of Energy’s Biological and Environmental Research (BER) program, advances Low Dose Radiation (LDR) research through multidisciplinary efforts across seven key thrusts. This document focuses on Thrust 1, which centers on the creation of curated multimodal population health datasets and supports broader efforts within the LUCID program, including AI-based hypothesis generation, experimental design, and the study of LDR-induced health risks. Specifically, it describes the identification and cataloging of Thrust 1’s curated LDR datasets and biodata, emphasizing their critical role in supporting various research thrusts within the consortium, with potential applications in healthcare and public policy. In addition, the document includes an evaluation of three Large Language Models (LLMs)—GPT-4, SOLAR-10B, and Mixtral-8x7B—based on their ability to extract features from 25 LDR studies. The results indicate that GPT-4 performed the best, while Mixtral-8x7B demonstrated limited knowledge. Overall, this work advances understanding in radiation protection, risk assessment, and medical treatments, while providing valuable resources for researchers, educators, and policymakers.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗