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At least 1,117 records · Page 62
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DOE OSTI2024

Overview of SAVY-4000 Lifetime Extension Activities in Fiscal Year 2024

This report provides an overview of activities engaged in during the 2024 fiscal year in support of the SAVY-4000 (hereafter “SAVY”) lifetime extension. As the final full year prior to the submittal of the technical basis, efforts were made to assess the current understanding regarding SAVY degradation in service as well as any knowledge that provided a more substantive view of the challenges that SAVYs face in storage. Close coordination with the stakeholders (i.e. LANL program office) ensured alignment with expectations regarding timelines for the remainder of the activities supporting the technical basis document for the lifetime extension request.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Wireless System Development for GaN-Based Transmitters for Advanced Nuclear Reactor Environments

The market for consumer electronics is dominated by Si-based electronics. Many Si-based complementary metal–oxide–semiconductor (CMOS) devices are reaching their limitations in terms of switching frequencies, voltages, currents, and temperatures. Furthermore, these devices are widely known to be susceptible to ionizing radiation and neutrons. Advanced nuclear reactors, especially microreactors and small modular reactors, require wireless electronics that can survive larger neutron and gamma doses than Si CMOS. Because of their size, electronics will need to be placed closer to the reactor cores, and cabling capabilities may be limited. However, GaN high-electron-mobility transistors (HEMTs) have been shown to switch faster with higher gains, and they can achieve current and voltage limits beyond traditional Si CMOS. However, they have less market maturity and have undergone less testing. These GaN HEMTs have been shown to survive ionizing radiation with little effects, but the effects of neutron radiation remain widely unknown. In this report, a wireless electronic system capable of supporting multiple radiation-hardened (rad-hard) sensors is presented. The proposed transmission scheme was designed to be simple to increase the likelihood of success in the immature GaN HEMT process and in high-radiation environments. The transmission system was tested using software-defined radios to emulate the transmission scheme before the transmitter circuitry in the GaN process was developed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Final Design for Additional Thermal/Epithermal eXperiments (TEX) with Sodium Chloride Absorbers to Provide Validation Benchmarks for TerraPower

The first set of Thermal/Epithermal eXperiments (TEX) with chlorine absorbers (TEX-Cl) were executed in Q4FY24 and are in the process of being benchmarked for the ICSBEP. TEX-Cl builds upon the TEX-HEU baseline cases that were published in the 2022 International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. TEX-HEU, like TEX-Pu, was designed to be modular to allow for the incorporation of various absorbers and reflectors to test nuclear data and application case needs. For example, TEX-HEU with hafnium (TEX-Hf) utilizes hafnium plates as both absorbers and reflectors, depending on the tested configuration. A second set of chlorine experiments, dubbed More TEX-Cl, are laid out in this report to meet the needs of TerraPower for chlorine validation for their Molten Chloride Fast Reactor (MCFR) systems. TerraPower’s Molten Chloride Reactor Experiment (MCRE) and MCFR are fast molten salt reactors that utilize sodium chloride (NaCl) salt eutectics as the fuel and coolant. The MCRE eutectic is a mixture of NaCl and uranium trichloride (UCl 3 ). An abundant need for chlorine absorption validation has been expressed by multiple members of the community, including Y-12 (whose needs were addressed with the first set of experiments), LANL (whose needs were addressed with the Chlorine Worth Study (CWS)), TerraPower, institute de radioprotection et de sûreté nucléaire (IRSN), Savannah River Nuclear Solutions (SNRS), and others. Of the members who have expressed interest in this validation, most are interested in the fast neutron energy region, where the 35 Cl(n,p) reaction is most prominent. New 35 Cl(n,p) differential cross section measurements performed by LANL at LANCSE show substantial changes to the cross sections (Figure 1) and may be validated through these experiments as some configurations are optimally sensitive to this cross section.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

High Energy Arcing Fault (HEAF) Photometrics 2022 Test Report

High Energy Arcing Faults (HEAFs) are hazardous events in which an electrical arc leads to the rapid release of energy in the form of heat, vaporized metal, and mechanical force. In Nuclear Power Plants, these events are often accompanied by loss of essential power and complicated shutdowns. To confirm the probabilistic risk analysis (PRA) methodology in NUREG/CR-6850, which was formulated based on limited observational data, the NRC led an international experimental campaign from 2014 to 2016. The results of these experiments uncovered an unexpected hazard posed by aluminum components in or near electrical equipment and the potential for unanalyzed equipment failures. Sandia National Laboratories (SNL), in support of the NRC work, collaborated with NIST, BSI, KEMA, and NRC to support the full-scale HEAF test campaign in 2022. SNL provided high speed visible and infrared video/data of ten tests that collected data from HEAFs originated on copper and aluminum buses inside switchgears and bus ducts. Part of the SNL scope was to place cameras with high-speed data collection at different vantage points within the test facility to provide NRC a more complete and granular view of the test events.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Non-invasive Pipe Pressure Monitoring for Safeguards

Our project explores a new solution to complement current safeguards toolbox by independently, and noninvasively, determining the pressure inside uranium hexafluoride (UF 6 )-carrying Gas Centrifuge Enrichment Plant (GCEP) pipes. Proof-of-principle, non-invasive pipe pressure monitoring tests were conducted through measurements of hoop and axial strain using an optical-interference dilatometry technique based on fiber Bragg gratings (FBGs) inscribed in the core of telecommunication-type 125 µm diameter optical fibers. The fibers were affixed to commercial stainless steel and aluminum tubes and pipes of the type expected to be found at GCEP’s, each measuring four inches in diameter. When the internal air pressure in the pipes changes their external diameter and length follow, and the pitch in the FBG shrinks and/or expands accordingly. In turn, these changes result in a shift of the light wavelength reflected by the FBGs. Several FBGs physically attached to the pipe, as well as control FBGs not attached, were interrogated during the process using state-of-the-art commercial interrogators that operate on swept-wavelength laser technology. Strains measured were contrasted against calculations based on the known elastic and mechanical properties of the pipes. Good agreement is found, and strategies are proposed for further improvement of the experimental resolution as well as future implementation of this novel approach for GCEP safeguards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

COG11.3 Abstract

COG is a high-resolution code for the Monte Carlo simulation of coupled particle transport in arbitrary3-D geometry. COG will transport neutrons, protons, deuterons, alpha particles with energies up to hundreds of GeV, and photons with energy ranges limited by the available cross section sets and physics models. Electrons can be transported via the EGS5 electron transport kernel, electrons can also be transported. The COG code is a significant upgrade from earlier Monte Carlo transport codes and has been written specifically to makes it more versatile, accurate, and easy to use. COG has provisions for calculating deep penetration (shielding) problems, criticality problems, and neutron activation problems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Benchmarking of Probabilistic Fracture Mechanics Models in Grizzly

In 2020, as part of contract/task order 31310019N0006/31310020F0060, the Nuclear Regulatory Commission tasked Idaho National Laboratory (INL) with benchmarking the MOOSE/GRIZZLY code against the NRC’s Fracture Analysis of Vessels - Oak Ridge (FAVOR), version 16.1 code. Both codes can predict large light-water reactor vessel integrity deterministically and probabilistically. In fact, such predictions are the primary purpose of the FAVOR code, and are a main focus of the GRIZZLY code, although GRIZZLY is capable of modeling a wider range of geometries and problems. The vessel integrity modeling capabilities of the GRIZZLY code are still under development and being gradually expanded, and the capability to model vessel fracture by crack growth has not yet been integrated into GRIZZLY (although it does exist in pre-release versions). Thus, the present benchmark study focused on comparing predictions for the conditional probability of crack growth initiation (CPI) between GRIZZLY and FAVOR v16.1. The present benchmarking study expands on a previous study completed in 2020, which focused on single plate regions, various flaw types, and full vessel studies. The previous study was limited in scope and revealed some relatively large discrepancies, particularly in the case of weld regions. In the present study, detailed investigations into the sources of discrepancies between GRIZZLY and FAVOR were performed, with a focus on weld regions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Hardware-Based Demonstration of Temperature Control Functions for Reactor Systems

Establishing autonomy in reactor control systems has become essential for the expansion of nuclear technologies. Thermal regulation in particular remains crucial for maintaining stable operation and ensuring the integrity of fuel. To alleviate public skepticism of the safety of nuclear reactors, demonstrating control over this key factor is pivotal. Utilizing electric heat pads to simulate the heat released in a reactor core, thermocouples for temperature monitoring, and an Arduino micro programmable logic controller (PLC) for control, a hardware-based demonstration of a reactor heating system validates the efficacy of reactor control over this key parameter. To improve precision, a proportional-integral-derivative (PID) algorithm was implemented in the heating control loop to ensure meticulous control of reactor functions. In addition, the integration of this physical system with a digital simulator tool such as RELAP5-3D establishes a foundation for a comprehensive testing environment. This allows for a refinement of temperature control under various simulated reactor conditions, bringing another layer of reliability to the operation of the system. By facilitating a physical demonstration of reactor thermal management and control strategies, this project provides a foundation for expanded testing and educational outreach. Ultimately, this system advances the broader goal of demonstrating the safety and viability of autonomous reactor operations, contributing to public trust and future reactor deployment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Computational Modeling of Graphite Degradation due to Molten Salt Infiltration and Wear

Molten-salt reactors (MSRs) represent a promising next-generation reactor design, with graphite serving as a moderator and/or reflector in several designs. However, due to limited experimental data and operational experience, a technical understanding of the structural integrity of graphite in molten salt environments remains incomplete. This report presents a modeling-based evaluation of graphite degradation in MSR environments, focusing on the effects of salt infiltration in fuel salt-based designs and surface wear in pebble bed reactor designs. The objective of this study is to enhance understanding of the structural integrity challenges posed by these degradation mechanisms and to provide a framework for assessing graphite behavior in MSRs. The first part of the report investigates the phenomenon of molten salt infiltration into graphite. This infiltration occurs when molten salt permeates the interconnected pore structure of the graphite moderator, driven by factors such as pressure differentials and the physical properties of both the salt and graphite. The infiltration process is influenced by characteristics of the pore structure, viscosity of the molten salt, and the interfacial energies between the graphite, salt, and the atmosphere within the graphite pore. Utilizing a coupled multiphysics modeling approach with Grizzly software, the study evaluates the stress induced by internal heat sources due to infiltration, which can lead to structural concerns. This evaluation is crucial for understanding how infiltration affects the mechanical integrity of graphite components in MSRs. The study considers the Molten-Salt Reactor Experiment (MSRE) graphite stringer geometry due to the availability of relevant data. Through detailed finite element analysis, the study examines stress distributions at varying infiltration percentages, revealing that stress levels increase with higher amounts of infiltration. Rare-event simulations, using the parallel subset simulation (PSS) framework, further quantify the failure probabilities under input uncertainties, with a user-specified failure metric. The PSS framework also identifies critical input parameters that significantly affect the stress values, including infiltration amount, thermal conductivity, and power density. Additionally, considering realistic reactor scenarios, the analysis was performed to account for the combined effects of radiation and infiltration, and modeling strategies on how to analyze new reactor designs or new graphite grades are discussed. The second part of the report focuses on wear mechanisms in pebble bed-based MSRs. As graphite fuel pebbles interact with the graphite reflector block, wear can result in material loss and the formation of surface defects, which may act as stress concentrators. A similar multiphysics modeling framework is employed to assess the impact of wear on the structural integrity of graphite components. This study considers a generic fluoride-cooled high-temperature reactor (gFHR) design due to the availability of comprehensive data. Worst-case scenario dimensions of the reflector blocks were analyzed under thermal and radiation conditions. Subsequently, wear in the form of idealized pits and grooves is modeled on the inner surface of the graphite block, with the maximum stress from previous simulations. The simulations show that groove-type defects are more detrimental than pits, leading to higher stress concentrations. Considering worst-case simulation scenarios and experimental wear rates, it was determined that the formation of a surface defect critical enough to affect the stress may not be possible in a gFHR design. Overall, the findings of this research contribute to the development of robust modeling tools for predicting graphite behavior under various operational conditions in MSRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

FY2025 Status Report: Model 9975 O Ring Fixture Long-Term Leak Performance

Leak testing experiments to monitor the aging performance of Viton® GLT and GLT-S O-rings used in the model 9975 shipping package have been ongoing since 2004 at Savannah River National Laboratory. Seventy tests using mock-up 9975 primary containment vessels (PCVs) with GLT O-rings were assembled and heated to temperatures ranging from 200 to 450 °F. Due to material substitution, fourteen tests with GLT-S O-rings were initiated in 2008 and heated to temperatures ranging from 200 to 400 °F. The conditioning temperatures are elevated compared to the calculated maximum O-ring temperature in a 9975 package in storage, 158 °F, to accelerate aging and enable observations of O-ring failures in a reasonable time frame. The mock-up PCV fixtures are leak tested periodically, and all GLT O-ring fixtures aged at 350 °F or above have failed to maintain a leak-tight seal. Eight GLT O-ring fixtures aged at 300 °F have failed after 2.8 to 5.7 years at temperature while the remaining fixtures at 300 °F were retired from testing following more than five years of aging without failure. Two of these retired fixtures were returned to testing and heated to 350 °F to evaluate the impact of additional heating at higher temperature for aged O-rings. Fixture #20 failed after 3 months while fixture #18 failed after 9 months at 350 °F. These O-rings demonstrated that aged and in-service O-rings can continue to be used, even at higher temperatures, after being in storage, and their leak performance are consistent with other samples at 350 °F. There has been one GLT O-ring fixture which failed after 13.4 years of aging at 200 °F. However, 20 other GLT O-rings aging at 200 °F have remained leak-tight for over 16.9 years and remain in test. There are two GLT O-ring fixtures at 270 °F; one fixture has failed after 12.9 years while the other fixture remains in test after 12.5 years. All GLT-S O-ring fixtures aged at 300 °F or above have failed their leak test. No failures have yet been observed in GLT-S O-ring fixtures aging at 250 °F for 14.9 years, while one GLT-S O-ring fixture failed after 12.4 years at 200 °F. The leak testing data to date suggest the GLT and GLT-S O-rings aging in the K-Area Complex (KAC) storage at temperatures of 158 °F might maintain a leak-tight seal for up to 59 years. Data from the O-ring fixtures are generally consistent with results from compression stress-relaxation testing and provide confidence in the predictive models based on those results. However, uncertainty exists in extrapolating these elevated temperature results to the lower temperatures of interest for normal storage in KAC. The collective data from these test efforts suggest the minimum O-ring service life at KAC normal storage conditions should be at least 34 years for GLT and GLT-S O-rings. Measurement of compression set in O-rings removed from failed fixtures, compared to that from KAC surveillance O-rings, indicate significant margin remains for O-rings still in service in 9975 packages in KAC. Aging and periodic leak testing will continue for the remaining 24 mock-up PCV fixtures.

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

Design, Construction, and Installation of In-Glovebox Pumped Actinide Molten Salt Loops

This report summarizes the work conducted in FY25 to construct and install pumped actinide-bearing molten salt loopsin a radiological glovebox. The establishment of such a capability will provide a platform in which technologies developed for Gen. IV molten salt reactors may be tested under relevant operating conditions. A review of convective molten salt systems is provided to frame Argonne’s pumped actinide loop relative to contemporary and historical systems. Moreover, the infrastructure needed to operate the system is detailed to provide additional context as to design metrics of the loops and supporting equipment. The pumps’ performance in water testing is outlined and used to inform operational activities planned in FY26. Finally, electrochemical experiments investigating fundamental reaction mechanisms governing reactive oxygen species that promote corrosion of molten salt media is also reported. In total, these combined loop installation and sensor development tasks will serve to aid in the future commercial deployment of molten salt reactors in the U.S.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Solving the Pulsed-Power Driven Inertial Fusion Energy Standoff Problem (Abbreviated Final Report)

The standoff problem in pulsed-power driven IFE (Inertial Fusion Energy) refers to the problem of electrically connecting a fusion target with the driver in a way that preserves the driver/target interface in the presence of a large fusion energy release (100’s of megajoules to 1000 megajoules). High yield fusion drivers for stockpile stewardship applications have similar concerns, but without the complication of the high repetition rate needed for energy production. All proposed IFE or high yield systems have the challenge of protecting the reactor vessel and driver from the energy release, but pulsed power drivers have the additional challenge of establishing an electrical connection in a way that does not create excessive debris or require costly/massive structures that must be expendable. Fortunately, there are conceptual solutions in the form of very low mass transmission lines, usually in the form of wires or foils, since very little mass is needed to overcome the magnetic forces driving the transmission lines apart or provide a low resistance pathway given the short pulse duration of the driver. Several important aspects of the standoff problem were analyzed in the course of this LDRD project. Conceptual means of introducing the low mass transmission lines into the reactor chamber were proposed, including analysis of methods to retain the power flow gap between the electrodes in the presence of the chamber environment. The energy losses due to ohmic dissipation and magnetic acceleration of the low mass transmission lines were evaluated for candidate driver pulses and different transmission line masses, and the fluence of x-ray and neutron energy on the persistent power flow structures that must survive the pulse of energy were evaluated. Finally, means of rapidly pumping down the chamber were explored as a way to return the system to low pressure following the multi-atmosphere post-pulse pressure rise. The overall conclusion from the analysis is that, while the reactor environment creates stressful conditions for a low mass transmission line standoff system, there are concepts expending 10 kg or less of electrode mass each pulse with the potential to mitigate the stresses and successfully drive fusion targets. Whether such a system could lead to a practical and economical fusion reactor would require extensive research and development beyond the scope of the feasibility study.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Advancing Multi-Hazard Risk and Safety Considerations for Aging Nuclear Facilities (Revision 1)

This project will demonstrate a multi-hazard time-dependent probabilistic risk assessment (PRA) approach for nuclear facilities considering aging-related deterioration of structures. A generic pressurized water reactor (PWR) reactor subjected to seismic mainshock-aftershock sequences considering the aging of the containment structure will be used as a case study to demonstrate the multi-hazard PRA approach. Using advanced modeling and simulation, seismic mainshock-aftershock fragility functions will be simulated for the containment structure considering aging effects. A multi-hazard PRA model for a generic PWR reactor will be built to quantify the multi-hazard core damage frequency (CDF) and large early release frequency (LERF) with explicit time-dependent modeling of event sequences. To date, the cascading impacts of multi-hazards are not adequately accounted for in the PRA models for nuclear facilities. In addition, for both initial and periodic evaluation of facilities to withstand natural phenomena hazards (NPH), deterioration of the SSCs due to aging and other effects is not adequately considered. By advancing multi-hazard considerations accounting for aging effects, this project will contribute to improved understanding of the safety of aging nuclear facilities. Project outcomes such as the multi-hazard CDF and LERF will also allow facility owners to optimize upgrade and retrofit protocols. This project is divided into two components: (1) advanced modeling and simulations; and (2) multi-hazard PRA. For the first component, Idaho National Laboratory’s (INL) Multi-hazard Analysis for STOchastic time-DOmaiN phenomena (MASTODON) and BlackBear codes will be used to simulate the seismic response and damage of a containment structure under cascading mainshock-aftershock sequences with aging effects. Uncertainties related to the seismic inputs, material parameters, and environmental factors such as temperature and humidity will be identified and propagated to the fragility functions. These fragility functions, which consider aging effects, will be time dependent. The second component will take the fragility information from the first component to build a multi-hazard time-dependent PRA model starting from a generic PWR model to evaluate the multi-hazard CDF and LERF and characterize the associated consequences. For this component, OpenPRA Web Application and SAPHIRE code will be used. Events such as the Fukushima Daiichi accident have highlighted the importance of considering the cascading impacts of multi-hazards for PRA. Moreover, many of the reactors in the current nuclear fleet in the United States (US) are already operating well beyond their initially planned design life, and applications for further extensions to operating licenses are being considered. Therefore, considering multi-hazard effects and aging deterioration in the NPH risk assessment process will contribute to the safety of both existing and future nuclear facilities. Outcomes of this project will thus directly benefit the standards DOE-STD-1020-16 and DOE-HDBK-1224-18.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Stockpile Stewardship and Nuclear Testing: A Technical Assessment

Since 1992, the United States has retained confidence in its nuclear weapon stockpile without performing any tests that produce nuclear yield. Instead, it has invested in a stockpile stewardship program (SSP) based on the same approach that was validated through fifty years of experience during the nuclear testing era. As the test moratorium continues and new information becomes available, it is both necessary and appropriate to periodically revisit the question of whether or not nuclear testing should resume. In a recent article by Dr. Mark Schneider, he asserts that the United States should return to nuclear testing to address the issues that negatively affect our country’s nuclear deterrent. A serious, technically informed evaluation of the most significant of the issues Dr. Schneider has raised is provided in this paper. The conclusion of this evaluation is that the strategy of a well-funded SSP coupled with rigorous assessments to identify if there is a specific need for a nuclear test is sound and cost-effective approach. Any decision otherwise needs to take into careful consideration all the classified SSP successes and data pertaining to the particular issues in question.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2025

IER-620 Pulsed-Neutron Die-Away (PNDA) with Oil Targets, CED3a

The Pulsed-Neutron Die-Away (PNDA) Testbed has previously been used to perform thermal scattering law validation measurements on HDPE and Lucite (IER-501) and water (IER-580). This experimental campaign expands the tests to include three additional hydrogenous materials relevant to various machining operations with fissionable materials. As such, it will use much of the same processes and equipment from the previous PNDA experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Final Design for Thermal/Epithermal eXperiments (TEX) with Lithium Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility (IER 575, CED-2 Report)

One of the main goals of the Thermal/Epithermal eXperiments (TEX) project is to use existing Nuclear Criticality Safety Program (NCSP) assets to create critical experiment plutonium and uranium test beds for materials important to criticality safety that have insufficient benchmark evaluations. The plutonium test bed experiments were completed in 2018 and are published in the 2020 edition of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The uranium test bed assemblies were completed in 2023 and accepted in the 2024 edition of the ICSBEP Handbook.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Pulsed-Neutron Die-Away Response of H 2 O Targets to a D-T Neutron Generator Pulse

Pulsed-neutron die-away (PNDA) experiments were completed at Lawrence Livermore National Laboratory (LLNL). The goal of these experiments was to provide a benchmark to validate thermal neutron scattering laws of H 2 O. The experiment was conducted with a deuterium-tritium (D-T) neutron generator producing pulses of 14.1 MeV neutrons that impinged on a moderating target.

3He Detectors↗