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DOE OSTI2026

Development of a Test-Bed for Testing and Refining EarthEn’s Supercritical CO 2 Based Energy Storage System

EarthEn’s energy storage concept leverages supercritical carbon dioxide (sCO 2 ) as a working fluid and relies on compact, high-performance components operating at elevated pressures and temperatures. To accelerate component development and reduce technical risk prior to larger-scale demonstrations, Oak Ridge National Laboratory (ORNL) developed a 100 kW-scale sCO 2 test-bed under a Cooperative Research and Development Agreement with EarthEn (CRADA NO. NFE-24-10050). The objective of the work was to design and construct a flexible experimental facility capable of reproducing key thermodynamic state points and heat-transfer conditions relevant to EarthEn’s thermal energy storage (TES) cycle, with particular emphasis on enabling development and evaluation of next-generation heat exchangers and TES concepts. The test-bed consists of a closed-loop sCO 2 circulation system housed within an open-topped enclosure. In its as-installed configuration, dense-phase sCO 2 is recirculated through a printed circuit recuperator, an electrically heated section, a throttling device used to simulate turbine expansion, and a water-cooled printed circuit heat exchanger that rejects heat to the building chilled-water system before returning to the pump. The pump is driven by a variable frequency drive, enabling controlled adjustment of flow and operating point. A comprehensive instrumentation suite was integrated to support both safe operation and high-quality data collection. Installed sensors include Coriolis flow meters for sCO 2 flow rate and density, resistance temperature detectors and thermocouples distributed throughout the loop (including the heated section and key heat exchanger ports), and pressure transducers for absolute and differential pressure measurements. The facility was designed to support high-pressure (19 MPa nominal) and high-temperature (575°C nominal) operation with credited overpressure protection provided by a rupture disk. Nominal operating conditions were selected to support 100 kW-class testing while maintaining flexibility for non-heated and heated shakedown, control development, and future integration of advanced TES test sections. In parallel with facility development, a system-level thermal-hydraulic model was created using Modelica-based tools to support component sizing, anticipate performance over targeted test conditions, and establish a framework for future model calibration against experimental data. At the conclusion of the project performance period, the facility was in final assembly, and the pressure boundary was nearly completed. However, several practical challenges associated with high-pressure/high-temperature systems and specialized component procurement impacted schedule and prevented initial pump-driven operation and full commissioning within the available resources. This report documents the as-built design, operating capabilities, and instrumentation, and it summarizes key lessons learned related to heater fabrication and testing, first-of-a-kind assembly factors, specialty flange supply constraints, and fill pump corrective actions. Finally, it outlines a phased plan for future commissioning and experimental campaigns, including control and instrumentation shakedown, heater characterization, model calibration, and testing at state points representative of EarthEn’s TES cycle.

25 ENERGY STORAGE↗

DOE OSTI2022

Prototype hardware design and testing of the global common module for the global trigger subsystem of the ATLAS phase-II upgrade

We report the High-Luminosity Large Hadron Collider (HL-LHC) will deliver more than ten times the integrated luminosity of the previous runs combined. Meeting its stricter throughput requirements poses new challenges to the Trigger and Data Acquisition (TDAQ) systems of the LHC experiments. Introduced in the framework of the ATLAS experiment’s HL upgrade, the Global Trigger (GT) is a new subsystem which will perform offline-like algorithms on full-granularity calorimeter data. The implementation of the GT’s functionality is firmware-focused and is composed of three layers: multiplexing (or data aggregating), global event processing, and demultiplexing interface to the central trigger processor. Each layer will be composed of several, similar nodes, hosted on replicas of identical hardware, the Global Common Module (GCM), an ATCA front board which is designed to be adopted throughout the entire GT subsystem. This article proceeds from the TWEPP 2021 conference and presents the GCM hardware design, performed in 2020, and focuses on some key results of its extensive testing performed in 2021.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

An Initial Assessment of the Design Margins of Different ASME Section III, Division 5 Design Rules

This report develops a method for assessing the design margin of the ASME Section III, Division 5 Class A design rules, focusing on a definition of margin as the ratio between the actual, expected component life and the ASME design life. Full inelastic finite element simulations with a complete damage model capturing the available creep, creep-fatigue, fatigue, and tension test failure data produce the expected component lives, which can then be compared to corresponding ASME design calculations. The focus of this particular work is on Alloy 617 and the ASME creep-fatigue design rules, but the approach is general and could be applied to other design limits and to other materials. A margin assessment of a representative Alloy 617 component shows the ASME creep-fatigue design rules applied to this component have a margin of 10 on design life. The report describes several challenges in completing a more comprehensive margin assessment of the ASME rules, in particular challenges in developing probabilistic assessment methods for high temperature structural components. The report details these challenge and discusses the immediate possibilities of the deterministic margin assessment method developed here.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

30 cm Drop Tests

The data from the multi-modal transportation test conducted in 2017 demonstrated that the inputs from the shock events during all transport modes (truck, rail, and ship) were amplified from the cask to the spent commercial nuclear fuel surrogate assemblies. These data do not support common assumption that the cask content experiences the same accelerations as the cask itself. This was one of the motivations for conducting 30 cm drop tests. The goal of the 30 cm drop test is to measure accelerations and strains on the surrogate spent nuclear fuel assembly and to determine whether the fuel rods can maintain their integrity inside a transportation cask when dropped from a height of 30 cm. The 30 cm drop is the remaining NRC normal conditions of transportation regulatory requirement (10 CFR 71.71) for which there are no data on the actual surrogate fuel. Because the full-scale cask and impact limiters were not available (and their cost was prohibitive), it was proposed to achieve this goal by conducting three separate tests. This report describes the first two tests — the 30 cm drop test of the 1/3 scale cask (conducted in December 2018) and the 30 cm drop of the full-scale dummy assembly (conducted in June 2019). The dummy assembly represents the mass of a real spent nuclear fuel assembly. The third test (to be conducted in the spring of 2020) will be the 30 cm drop of the full-scale surrogate assembly. The surrogate assembly represents a real full-scale assembly in physical, material, and mechanical characteristics, as well as in mass.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Characterization of Hanford LAW phase 3 glasses. Revision 2

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Characterization of Hanford LAW Phase 3 Glasses

This report provides analyses of glass compositions, wash solution compositions, and Product Consistency Test (PCT) leachate compositions for a series of simulated low-activity waste (LAW) glasses fabricated at the Pacific Northwest National Laboratory (PNNL). These data will be used in the development of improved property/composition models for LAW glass at Hanford. Chemical analyses were performed on a representative sample of each of the quenched and sulfur saturated glasses to allow for comparisons with the targeted compositions. For some of the quenched glasses, measured concentrations of Cr₂O₃, K₂O, Na₂O, P₂O₅, and ZrO₂ were below the targeted values. For some of the sulfur saturated glasses, the measured concentrations of Al₂O₃, Cl⁻, Cr₂O₃, F⁻, K₂O, Na₂O, P₂O₅, V₂O₅, and ZrO₂ were generally low relative to the targeted values. A comparison of the measured compositions of the quenched and sulfur saturated versions of the study glasses showed that the measured concentrations of Al₂O₃, Cl⁻, and F⁻ were lower for most of the sulfur saturated glasses. The measured Cr₂O₃ and K₂O concentrations were markedly lower for the sulfur saturated versions of the glasses as compared to the quenched versions, which may indicate that these elements partitioned to the soluble sulfur salts. A set of modified glasses was received from PNNL later in this study. Chemical analysis showed that these glasses generally met their targeted compositions. The measured SO₃ concentrations were higher for most of the sulfur saturated glasses relative to those of the quenched versions, as expected. Chemical analysis of the wash solutions that resulted from preparation of the sulfur saturated melts identified Cr, K, Na, P, S, and V as the major elements in solution. With the exception of sulfur, this may explain why the measured concentrations of these components in the study glasses were generally below the targeted values. PNNL performed PCTs on quenched and canister centerline cooled (CCC) versions of the study glasses, as well as the modified glasses. The leachates were sent to SRNL for chemical analysis. Minor scatter in the measured values among the triplicate samples for each glass was noted. Several of the study glasses, both quenched and CCC, have NCi values that are greater than the immobilized LAW constraint of 4 g/L for B, Na, and Si.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

FY2021 Rotating Bed Reactor for Iodine Removal from ETF Feed to Support DFLAW Start-Up

In FY20, Washington River Protection Solutions, LLC (WRPS) began a program of work to evaluate potential technologies for removal of iodine from the ETF feed. Work at the Pacific Northwest National Laboratory (PNNL) focused on ion exchange media and conventional column based deployment while Atkins and VSL focused on ion exchange media in a Rotating Bed Reactor (RBR) system. In the RBR system, the ion exchange media is loaded into a packed bed, which is rotated in the liquid to be treated. The work described in this report builds on the results from FY20 work, in particular, with regard to improving iodate removal, extending the data set to support process modeling, further comparison of RBR versus column performance, and RBR scale-up testing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Corrective Action Unit 98: Frenchman Flat Five-Year Evaluation Nevada National Security Site, Nevada (Revision 1)

The Corrective Action Unit (CAU) 98, Frenchman Flat (FF) Closure Report (CR) (NNSA/NFO, 2016) and its ensuing Records of Technical Change (ROTCs) (compiled in DOE/EMNV, 2019) required the U.S. Department of Energy (DOE), Environmental Management (EM) Nevada Program’s Underground Test Area (UGTA) Activity to collect data for five years to establish triggers for the monitoring network and analyze the monitoring points for viability in long-term monitoring. This document provides the results of the five-year evaluation and presents recommendations for future monitoring of CAU 98.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2023

ENDF/B-VIII.1 Validation (Preliminary Release 2 + New 239 Pu)

This document should serve as a reference for validation testing of nuclear data files included in the second preliminary release of ENDF/B-VIII.1 and the newly evaluated 239 Pu nuclear data file to be included in the third preliminary release of ENDF/B-VIII.1. Experiment names listed in this report are based on International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook designations. Additional information on ICSBEP handbook designations can be found at https://www.oecd-nea.org/jcms/pl_20291/icsbep-handbook. The validation metrics used in this report include the effective neutron multiplication factor, k eff , and tritium production. All validation metrics were calculated using the radiation transport code Monte Carlo N-Particle (MCNP).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1

The fuel characterization models of SAS4A/SASSYS-1 (SAS) have recently been extended to include a new U-Pu-Zr metal fuel model, MFUEL. MFUEL is equipped with mechanistic physics based models to predict the pre-transient characterization and transient response of metal fuel, with emphasis on fuel melting, cladding failure, and the metal fuel’s impact on core reactivity. The MFUEL model will be available in the full version of SAS4A/SASSYS-1 5.7, which is scheduled to be released in June 2023. Fast reactor fuel pins that operated in EBR-II and FFTF with low smear density U-Zr and U-Pu-Zr metal fuels and irradiation resistant ferritic-martensitic cladding showed significant advantages in achieving high burnups and assuring inherent safety characteristics during anticipated transients, design basis events, and beyond design basis events. For safety analysis, a fuel performance model must be able to predict (1) Fuel pin mechanics and compositional and dimensional changes, (2) Clad failure, and (3) Fuel pin thermal resistance. Achieving these high level goals accurately is strongly related to the model performance of individual physical processes taking place within a fuel pin during its lifetime. Metal fuels typically operate above the mid-point temperature of melting during normal, as well as off-normal, conditions. At these elevated temperatures, the availability of thermal activation provides a driving force for various diffusional processes leading to complex phase transformations, micro-structure evolution, significant amounts of fuel swelling, interconnected porosity formation, excessive amounts of fission gas release, and fuel clad chemical interactions. Clad failure in fast reactors primarily occurs as a result of creep rupture augmented by clad wastage formation. The reaction is driven by thermal creep induced dislocation motion, grain boundary cavity nucleation, growth and breakup of grain boundaries. The high level complexity and limited available data requires introducing physics-based modeling approaches to gain extrapolation ability and sensitivity with respect to various conditions. The objective of this report is to perform validation of MFUEL using the experimental data for (1) Normal operation EBR-II fuel behavior, (2) Normal operation PHENIX fuel behavior, (3) HT9 Pressure tube ramp-and-hold creep rupture tests, (4) Whole Pin Furnace (WPF) creep strain, creep rupture and eutectic tests, (5) Fuel Behavior Test Apparatus (FBTA) eutectic tests, and (6) TREAT M5-7 OverPower tests up to clad failure. Section-2 includes a brief description of the MFUEL models. A detailed description of the MFUEL physics-based, semi-empirical models will be presented in the SAS V 5.7 theory manual. Section-3, 4, and 5 describes the validation effort for the pre-transient irradiation, furnace transients, and TREAT M-Series transients, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Rapid Demonstration of Bremsstrahlung Diode Optimization

Optimization of the radiation pattern from a Bremsstrahlung target for a given application is possible by controlling the electron beam that impacts the high-atomic-number target. In this work, the electron beam is generated by a 13MV vacuum diode that terminates a coaxial magnetically insulted transmission line (MITL) on the HERMES-III machine at Sandia National Labs. Work by Sanford introduced a geometry for vacuum diodes that can control the flow within bounds. The "indented anode", as coined by Sanford, can straighten out the electron beam in a high-current diode that would otherwise be prone to beam pinching. A straighter beam will produce a more forwardly directed radiation pattern while a pinching electron beam will yield a focal point or hot spot on axis and a more diffuse radiation pattern. Either one of these may be desirable depending on the application. This work serves as a first attempt to optimize the radiation pattern in the former sense of collimating the radiation pattern given a limited parameter space. The optimization is attempted first using electromagnetic particle-in-cell simulations in the EMPIRE code suite. The setup of the models used in EMPIRE is discussed along with some basic theory behind some of the models used in the simulations such as anode heating and secondary ions. Theoretical work performed by Allen Garner and his students at Purdue is included here, which concerns the impact of collisions in these vacuum diodes. The EMPIRE simulations consider both an aggressive and a conservative design. The aggressive design is inherently riskier while the conservative design is chosen as something that, while still a risk, is more likely to perform as expected. The ultimate goal of this work was to validate the EMPIRE code results with experimental data. While the experiment that tested the diode designs proposed by the simulation results fell outside of the fiscal boundaries of this project (and for that reason the results of which are not included in this report), the hardware for the experiment was designed and drafted within those same fiscal boundaries, and is thus included in this report. However, there was yet another experiment performed in this project that tested a key feature of the diode: the hemispherical cathode. Those results are documented here as well, which show that the cathode tip is an important aspect to controlling the diode flow. A short series of simulations on this diode were also performed after the experiment in order to gain a better understanding of the effect of ions. on the flow pattern and faceplate dose profile.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

Visual Style Elements in Human–System Interface Design for Nuclear Power Operations: Does Style Affect Performance and Preference?

As the world increasingly adopts renewable and sustainable energy systems, transitionary solutions include nuclear power, which currently provides 20% of the United States’ electricity and is the largest single source of carbon-free electricity generation. Advanced reactors are a critical component of a carbon-free mixed energy portfolio that require careful design of first-of-a-kind control rooms. The application of Human Factors Engineering (HFE) is essential for scientific and iterative testing of novel human–system interface (HSI) concepts to ensure effective, efficient, and safe plant operations. Microworlds are simulators that use simplified physics models and control systems to distill nuclear power operations into essential functions. HFE scientists used the Rancor Microworld Simulator to obtain preference and performance metrics for novel and traditional static HSI design styles. Participants comprised advanced reactor company employees and nuclear industry consultants. A mixture of quantitative and qualitative data was captured. There was a preference for the basic graphical style that included high contrast and traditional color scheme elements. No single HSI design outperformed the others, and the participants did not perform better using their preferred HSI style. We report this experiment is the first in a series of HFE testing for HSIs in advanced reactor control room development. Clear user preferences emerged for elements within static displays. The cutting-edge neumorphic style was the least preferred. Future directions include tests of dynamic displays. HFE is used in evaluating and designing HSI devices that will improve the efficiency and safety of advanced nuclear power operations.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2025

Comparative post-irradiation examination of high burnup U-19Pu-10Zr: Assessing steady-state irradiation behavior against historical and modeled fuel performance

Here, the development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.

BISON↗

DOE OSTI2024

Measurements of radioxenon activities during periods of gaseous release from an advanced gas-cooled reactor

Activities of radioxenon isotopes are reported from an Advanced Gas-cooled Reactor (AGR) during periods of gaseous release. Xenon isotopes are relevant to the monitoring regime of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). These releases may affect detections from the International Monitoring System (IMS) by influencing the radioxenon background. Time series activity plots have been produced from both in-core monitoring and direct measurement at the point of release using a stack monitor system. Ratio plots using both data sets have been produced with results compared with a commonly chosen nuclear explosion “discrimination line.” The reported results improve the understanding of radionuclide emissions from civil nuclear power plants. This work is being performed as part of the Xenon Environmental Nuclide Analysis at Hartlepool (XENAH) collaboration between the Atomic Weapons Establishment (AWE, UK), EDF Energy (UK), Pacific Northwest National Laboratory (PNNL, USA), and the Swedish Defence Research Agency (FOI).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Galaxy zoo builder: Morphological dependence of spiral galaxy pitch angle

ABSTRACT Spiral structure is ubiquitous in the Universe, and the pitch angle of arms in spiral galaxies provide an important observable in efforts to discriminate between different mechanisms of spiral arm formation and evolution. In this paper, we present a hierarchical Bayesian approach to galaxy pitch angle determination, using spiral arm data obtained through the Galaxy Builder citizen science project. We present a new approach to deal with the large variations in pitch angle between different arms in a single galaxy, which obtains full posterior distributions on parameters. We make use of our pitch angles to examine previously reported links between bulge and bar strength and pitch angle, finding no correlation in our data (with a caveat that we use observational proxies for both bulge size and bar strength which differ from other work). We test a recent model for spiral arm winding, which predicts uniformity of the cotangent of pitch angle between some unknown upper and lower limits, finding our observations are consistent with this model of transient and recurrent spiral pitch angle as long as the pitch angle at which most winding spirals dissipate or disappear is larger than 10°.

Lingard, Timothy↗

DOE OSTI2022

Utah FORGE Well 16A(78)-32 Stimulation Data (April, 2022)

This is a set of data related to the stimulation program at Utah FORGE well 16A(78)-32 during April, 2022. This includes daily reports, 1 second Pason data, tracer data, and shear stimulation data and information including a report of an evolving prognosis for the stimulation operations.

15 GEOTHERMAL ENERGY↗

DOE OSTI2021

Attribute-Guided Adversarial Training for Robustness to Natural Perturbations

We report while existing work in robust deep learning has focused on small pixel-level norm-based perturbations, this may not account for perturbations encountered in several real world settings. In many such cases although test data might not be available, broad specifications about the types of perturbations (such as an unknown degree of rotation) may be known. We consider a setup where robustness is expected over an unseen test domain that is not i.i.d. but deviates from the training domain. While this deviation may not be exactly known, its broad characterization is specified a priori, in terms of attributes. We propose an adversarial training approach which learns to generate new samples so as to maximize exposure of the classifier to the attributes-space, without having access to the data from the test domain. Our adversarial training solves a min-max optimization problem, with the inner maximization generating adversarial perturbations, and the outer minimization finding model parameters by optimizing the loss on adversarial perturbations generated from the inner maximization. We demonstrate the applicability of our approach on three types of naturally occurring perturbations --- object-related shifts, geometric transformations, and common image corruptions. Our approach enables deep neural networks to be robust against a wide range of naturally occurring perturbations. We demonstrate the usefulness of the proposed approach by showing the robustness gains of deep neural networks trained using our adversarial training on MNIST, CIFAR-10, and a new variant of the CLEVR dataset.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2020

Sister Rod Destructive Examinations (FY20) Appendix B: Segmentation, Defueling, Metallograhic Data and Total Cladding Hydrogen

As a part of the US DOE-NE High Burnup Spent Fuel Data Project, ORNL is performing destructive examinations (DEs) of high burnup (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the “sister rods,” are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO®, and M5®. The DEs are being conducted to obtain a baseline of the HBU rod’s condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding are being tested. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities.This report documents the status of the ORNL Phase 1 DE activities related to: rough segmentation, defueling, optical microscopy, and cladding total hydrogen measurements for 7 Phase 1 sister rods and outlines the DE tasks performed and the data collected to date. The results of these detailed examinations and others are summarized in the Sister Rod DE Status Report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗