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TRISO development background and particle design US DOE AGR Program Fuel fabrication and quality control Irradiation testing Post-irradiation examination and high-temperature safety testing
Oak Ridge National Laboratory (ORNL) was a pioneer in the testing of radioactive material shipping packages and has been performing such tests for over 50 years. Currently, except for Type B thermal tests, testing is performed at the purpose-built Packaging Research Facility that is located at the National Transportation Research Center - a U.S. Department of Energy User Facility. Packages used for the transportation of Type B quantities of radioactive materials must be capable of meeting normal conditions of transportation (NCT) requirements and hypothetical accident conditions (HAC) requirements as defined in Title 10 Code of Federal Regulations (CFR) Part 71. Evaluation of package designs to these criteria typically involves a combination of analysis and physical testing. Required physical tests include free drop, crush, puncture, penetration, compression, vibration, water spray, water immersion, and thermal. The Oak Ridge National Laboratory (ORNL) Packaging Evaluation Facility (PEF) at the National Transportation Research Center (NTRC) provides an inclusive testing capability for a wide variety of packages needed for shipment of radioactive materials. Since PATRAM 2019, ORNL has performed several sets of tests including testing of two designs of the Defense Program Package (DPP-3 and DPP-1) and several Special Form Capsules. This paper provides an overview of the testing performed as well as a summary of advancements made to the ORNL package testing program during that time.
There are good ideas and new technologies proposed every day to solve problems within the DOE complex. A process to transition a new technology from inception to the decision to launch a project with baselines is described. Examples from active technology development projects within Savannah River Remediation (SRR) will be used to illustrate the points. The process includes decision points at key junctures leading to preliminary design. At that point, normal project management tools can be employed. The technology development steps include proof-of-principle testing, scaled testing and analysis, and conceptual design. Tools are used that define the scope necessary for each step of technology development. The tools include use of the DOE technology readiness guide, Consolidated Hazards Analysis (CHA) and internal checklists developed by Savannah River Remediation. Integration with operating or planned facilities is also included. The result is a roadmap and spreadsheet that identifies each open question and how it may be answered. Performance criteria are developed that enable simple decisions to be made after the completion of each step. Conceptual design tasks should begin as the technology development continues. The most important conceptual design tasks at this point in the process include process flow diagrams (PFDs), high level Process and Instrumentation Drawings (P and IDs), and general layout drawings. These should influence the design of the scaled simulant testing. Mechanical and electrical drawings that support cost and schedule development should also be developed. An early safety control strategy developed from the CHA will also influence the cost. The combination of test results, calculations and early design output with rough order of magnitude cost and schedule information provide input into the decisions to proceed with a project and data to establish the baseline. This process can be used to mature any new technology, especially those that must be integrated into complicated flow sheets. Then fully informed decisions can be made to usher technology development ideas through project implementation. Examples from active technology development projects, Tank 48 Treatment and Enhanced Chemical Cleaning, are presented. This model for technology development integrates several tools and activities to provide a logical method for maturing new ideas and technologies. It provides a framework to develop a good idea into a full up design/build project. This model defines a set of integrated tasks to mature the technology and define scope, cost and schedule necessary to make business decisions for implementing technology development. (authors)
This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. The DOE Office of Nuclear Energy (DOE-NE) Office of Spent Fuel and Waste Disposition (SFWD) Integrated Waste Management System (IWMS) program is developing an Execution Strategy Analysis (ESA) Origin Sites Readiness Model using the Goldsim{sup TM} software application to provide information on potential future approaches for removing SNF and HLW from origin sites such as commercial nuclear power plant sites. The purpose of this ESA model is to apply integrated waste management system analysis, system engineering, and decision analysis principles to inform future decisions regarding potential future nuclear waste management system architectures. The ESA Origin Sites Readiness Model is a dynamic simulation modeling capability. It is developed for use in the analysis and comparison of potential future implementation strategies associated with an integrated nuclear waste management program. The model is designed to be independent of the destination of the SNF or HLW so the results from the Origin Sites Readiness Model are equally applicable to a wide range of scenarios involving shipments to generic destinations in an integrated waste management system. The model explicitly takes significant assumptions, uncertainties, and risks into account. Currently, the ESA Origin Sites Readiness Model includes 14 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Vermont Yankee, La Crosse, Big Rock Point, Zion, Kewaunee, Fort Calhoun, Trojan, Rancho Seco, Humboldt Bay, San Onofre, and Crystal River) and is designed to be modular so that other sites can be added in the future. The model also includes multiple transportation modes (heavy haul truck, rail, and barge), activities related to the acquisition of transportation assets such as railcars and transportation casks, and site-specific considerations of on-site and near-site transportation infrastructure. The model was developed using the ESA process with Subject Matter Experts (SME); it incorporates uncertainty and risks and provides a structured, systematic methodology for evaluating potential SNF campaign scenarios. The model also complements other system-level logistics tools being developed by SFWD. (authors)
NiTi and NiTi-based Shape Memory Alloys (SMA) exhibit a reversible solid-state phase transformation from martensite to austenite driven by thermal energy. High temperature (Mf>100°C) SMAs are martensite at room temperature and can be fabricated into solid-state actuators that return to a pre-programmed shape against a designed load after heating to transformation threshold. Reactive as-fabricated additively manufactured parts (4-D printing) is the current state of the art in manufacturing of SMAs but requires compositions compliant to rapid solidification. Existing actuator designs are developed from commercially available, highly investigated material compositions. However, existing high temperature high performance (high actuation strain, low thermal hysteresis) shape memory alloys contain significant (>10% at.) portions of high-cost Platinum Group Metals (PGMs). It is of significant scientific interest to investigate material compositions that are peer performing or superior to PGMs whose constituent elements represent a significant cost savings. Shape memory alloy properties vary significantly with small (0.1% at.) compositional changes making robust investigative sample sets very large. Computational material design can be deployed to shrink the compositional space of possible alloy combinations and reduce the experimental load in material discovery. Investigating shape memory effect (SME) and validating process additive process parameters for a single novel composition is cost intensive in both time and consumed materials. Additionally, sub-optimal processing, oxygen, or solidification rate sensitivity could render additively manufacturing specimens without micro, macro cracks, or significant chemical variance impossible. Unfortunately, such failure susceptibility cannot be simulated. Therefore, a research pathway to validate novel shape memory alloy compositions for powder bed fusion additive manufacturing without the need for powdered feedstock is also proposed. This research investigates novel high temperature shape memory alloys for actuators without platinum group alloying elements to discover one that could be commercially viable as an additive manufacturing feedstock.
To validate and verify the python-based code (PySL), designed to replicate the programs used by STAYSL for Beam Correction Factor (BCF) and Self-Shielding Factor (SHIELD), a series of tests were performed. To test BCF a python script was written to generate a random flux history file and both versions of the code processed the data. The test verified matching values up to at least one decimal place, approximately 10,000 tests where run and each one passed. Isotopes began to fail the tests once neutron saturation was reached. To verify this the total time of exposure was varied the isotopes that failed were compared to a list of their half-lives. The test process for SHIELD was very similar but, in this case, the code began by producing an input file with varying thickness and device type/environment for the SHIELD input. The failure condition for this test was if any of the data points for an isotope had a difference above 3%. Approximately 40 of these tests were run and there were only 3 isotopes that had reoccurring failures but only 2% of their points were above the 3% difference. A visual comparison was conducted by plotting the results from both programs. Although the test failed, the differences between their values were minuscule, and the self-shielding factor’s shape was preserved when plotted. Next steps for this project will be validating and verifying the python-based SigPhi code and then reproducing and testing the least squares unfolding performed by STAYSL.
SUMMARY The U.S. Department of Energy (DOE) Advanced Reactor Technologies (ART) Graphite Research and Development (GRD) Program is conducting a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. This report documents the qualification of irradiation monitoring data for the fifth experiment, High Dose Graphite-1 (HDG-1). Qualified monitoring data are required by the ART program to support the design and licensing of the first high-temperature reactor (HTR) nuclear plant. Data are classified as Qualified if they meet the usage requirements described in the experiment planning and quality assurance (QA) documents, Failed if they do not meet those requirements and provide no usable information, or Trend if they do not fully meet all requirements but still provide useful information subject to an assessment of how any deficiencies may affect a particular use of the data. HDG-1 irradiation began with Advanced Test Reactor (ATR) Cycle 168B on August 24, 2020, and concluded after Cycle 173C on January 27, 2025. The HDG-1 capsule was removed from the reactor core twice—during core internal change (CIC) Cycle 170A and powered axial locator mechanism (PALM) Cycle 172A—to prevent overheating of the graphite specimens during high-power PALM cycles. The capsule was therefore irradiated during a total of seven normal ATR cycles: 168B, 169A, 171A, 171B, 173A, 173B, and 173C. Irradiation monitoring data evaluated in this report include thermocouple (TC) temperature, gas flow rate, gas moisture, gas pressure, specimen load, and graphite stack displacement. Temperature. A total of 14,508,065 TC temperature records were captured. Of these, 13,901,785 (95.8%) are Qualified and 606,280 (4.2%) are Failed. The principal source of failed temperature data was the instrument failure of TC-9 (Zone 2) on June 24, 2024, and TC-10 (Zone 1) on July 5, 2024, near the end of Cycle 173A, which resulted in 595,554 Failed readings. An additional 379 missing values and 10,347 slightly negative values from TC-13 during ATR outages are also Failed. Neither TC-9 nor TC-10 was used as a temperature-control TC, and their failures did not compromise capsule condition monitoring. Correlation analysis of all 13 TCs found no evidence of virtual junction formation. Control chart analysis revealed clear downward drift of approximately 80°C for TC-6 (Zone 3) relative to other stable TCs, and possible downward drift of approximately 60°C for TC-13 relative to the Zone 5 control TC (TC-1), though TC-13 remained consistent with the Zone 2 control TC (TC-12). Gas flow. A total of 20,088,090 gas flow rate records were captured. Of these, 19,941,463 (99.3%) are Qualified and 146,627 (0.7%) are Failed due to missing values. All argon, helium, and total gas flow data were within expected ranges throughout the irradiation. Gas moisture. A total of 1,116,005 outlet gas moisture values were captured. Of these, 1,101,421 (98.7%) are Qualified and 14,584 (1.3%) are Failed, comprising 14,556 out-of-range values and 28 missing values. The out-of-range moisture values exceeded 22,000 ppmv for approximately 1 week at the beginning of Cycle 173A, when accumulated moisture evaporated after the capsule was retrieved from water storage during PALM Cycle 172A and reinserted into the east flux trap. Moisture levels returned to below 25 ppmv for the remaining three cycles, and the transient high-moisture event did not affect the integrity of specimen irradiation. Gas pressure. A total of 7,812,035 gas pressure values were captured. Of these, 6,642,048 (85.0%) are Qualified and 1,169,987 (15.0%) outlet pressure values are Failed, comprising 718,537 zero outlet pressure values due to sensor failure from Cycle 168B through Cycle 171B, 54,550 missing values, and 396,900 too-low outlet pressure values, ranging from 1.1 to 1.6 psia after sensor replacement during Cycle 173A. Load. A total of 6,696,030 load values were captured. Of these, 6,694,580 (99.98%) are Qualified and 1,450 (0.02%) are Failed due to missing values. Applied loads to the six specimen stacks were stable throughout the irradiation. Stack displacement. A total of 6,696,030 displacement values were captured. Of these, 5,713,297 (85.32%) are Qualified and 3,781 (0.06%) are Failed due to missing values. Stack displacement increased consistently throughout the irradiation, reaching approximately 3.08 in. for Channels 5 and 6 by the end of irradiation. 978,952 (14.62%) substantially elevated displacements observed for Channel 6 beginning in Cycle 171A and for Channel 5 beginning in Cycle 173A are assigned Trend status. Raising pressure. A total of 1,115,999 raising pressure values were captured. Of these, 1,115,430 (99.95%) are Qualified and 569 (0.05%) are Failed due to missing values. Ram pressure. A total of 6,696,030 ram pressure values were captured. Of these, 6,692,249 (99.95%) are Qualified and 3,484 (0.05%) are Failed due to missing values. Stack raising was perf
Idaho National Laboratory (INL) is developing a microreactor to produce electrical power utilizing a small nuclear core under the Microreactor Application Research Validation and Evaluation (MARVEL) program. Key to this design are neutron reflector components constructed of Beryllium Metal (S-65) which is a structural grade of hot-pressed beryllium block. MARVEL will be located at the Transient Reactor Test (TREAT) facility. Any conflict between this specification and referenced Codes and Standards, or any supplementary specifications in the procurement documents requires written clarification from the Contractor prior to proceeding with any work. Any deviation from the procurement documents requires approval by the Contractor with the change request process.
The Palisades Nuclear Generating Station included in its original surveillance program a surveillance capsule, designated A-60. The capsule was removed from its surveillance position in early 1995 and has been resident in the spent fuel pool since that time. It was harvested to perform characterization of surveillance specimens in this capsule in 2023. This capsule was irradiated to a fluence of 1.96x1020 n/cm2 (E> 1MeV) that is equivalent for more than 150 effective full power years for the current US reactor pressure vessel (RPV) fleet. This capsule contained several materials, including Charpy specimens of standard reference material (SRM) from Heavy-Section Steel Technology (HSST) A533-B Plate 01. To perform Charpy testing of this highly irradiated material, the new Charpy specimen transfer system was designed and implemented on Charpy impact testing machine in the hot cell to accommodate remote testing of these specimens. This new transfer system includes integrated environmental chamber to cool or heat Charpy specimens to the desired temperature. Testing of this highly irradiated material revealed very large shift of Charpy transition temperature, 188oC.
This document serves as the final report for the AFRL SBIR Program "Directionally-Tailored Infrared Emission and/or Transmission" AF 1 O-BT30 The focus of this program was in the development and design of a thermal emission surface with thermal emission directed away from broadside in the L WIR band. In the report, a basic overview is provided covering past research efforts in thermal emission surfaces and the fundamental science of directional emitters. Multiple sets of designs are investigated as candidate directional emission surfaces: dark element, image, and end-fire linear array, broadside emitter, and waveguide emitter. Based on program results, both the broadside emitter and waveguide emitter designs appear to meet the needs of the AF directional emission requirements. These designs offer a range of directional emission behavior (controllable emission beamwidth and bandwidth) and scale well to mass production. The designs require additional environmental testing, but appear to be suitable for space application as well. All goals of the program appear to have been met and the team is looking forward to future development.
The high performance computing (HPC) community is in a period of transition. The rise of AI/ML coupled with a changing landscape of resources deems portability a new metric of performance, and methods to move between on-premises and cloud environments and assess compatibility are paramount. Here we design and test a strategy for bridging the gap between traditional HPC and Kubernetes environments – first containerizing applications, providing automated orchestration to run studies, and packaging the setup with automated means to assess performance using low overhead eXtended Berkeley Packet Filter (eBPF) programs. We first assess different designs for eBPF collection, demonstrating a tradeoff between number of programs deployed on a node and overhead added. We develop 5 low overhead eBPF programs that combine with streaming ML models to assess CPU, futex, TCP, shared memory, and file access across four different builds of an HPC application for CPU and GPU. We use eBPF data to generate insights into the possible underlying etiology of scaling issues. We then assess compatibility of a well-known benchmark, HPCG, across matrices of micro-architectures and optimization levels (217 containers across 24 instance types and over 7500 runs). We provide to the community 30 applications to deploy in our automated setup and perform a scaling study from 4 to a maximum of 256 nodes for both CPU and GPU applications. Finally, we use our gained knowledge about performance to generate compatibility artifacts that are used by a newly developed Kubernetes controller to intelligently select instance type based on optimizing a figure of merit. Along with insights to scaling in this environment with a collection of applications and templates to work from, we provide an overall strategy for approaching HPC application deployment and image selection based on compatibility in cloud.
The Apache Point Observatory Galactic Evolution Experiment 2 (APOGEE-2) is a dual-hemisphere, near-infrared (NIR), spectroscopic survey with the goal of producing a chemodynamical mapping of the Milky Way. The targeting for APOGEE-2 is complex and has evolved with time. In this paper, we present the updates and additions to the initial targeting strategy for APOGEE-2N presented in Zasowski et al. (2017). These modifications come in two implementation modes: (i) “Ancillary Science Programs” competitively awarded to Sloan Digital Sky Survey IV PIs through proposal calls in 2015 and 2017 for the pursuit of new scientific avenues outside the main survey, and (ii) an effective 1.5 yr expansion of the survey, known as the Bright Time Extension (BTX), made possible through accrued efficiency gains over the first years of the APOGEE-2N project. For the 23 distinct ancillary programs, we provide descriptions of the scientific aims, target selection, and how to identify these targets within the APOGEE-2 sample. The BTX permitted changes to the main survey strategy, the inclusion of new programs in response to scientific discoveries or to exploit major new data sets not available at the outset of the survey design, and expansions of existing programs to enhance their scientific success and reach. After describing the motivations, implementation, and assessment of these programs, we also leave a summary of lessons learned from nearly a decade of APOGEE-1 and APOGEE-2 survey operations. Finally, a companion paper, F. Santana et al. (submitted; AAS29036), provides a complementary presentation of targeting modifications relevant to APOGEE-2 operations in the Southern Hemisphere.
Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion to ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.
The Transformational Challenge Reactor (TCR) program aims to demonstrate a revolutionary design approach enabled by advanced manufacturing and data analytics in the design of nuclear reactors. This article discusses scoping analyses of preconceptual designs to inform TCR design decisions and the evaluation of sensitivities and uncertainties on postulated transient scenarios. The applicability of the systems codes TRACE and RELAP5-3D to TCR transient analysis are examined, and RELAP5-3D models are used to examine the transient response of two candidate core designs at multiple power levels. Then, the uncertainty quantification code RAVEN is used to quantify the effect of several design parameters on reactivity-initiated accident (RIA) progression at hot zero power (HZP) and hot full power (HFP) as well as to assess the impact of uncertainties in transient parameters for the pressurized loss of forced cooling (PLOFC).When results were compared, TRACE and RELAP5-3D showed good agreement in their ability to predict system behavior, but RELAP5-3D calculations were closer to analytical predictions for the RIA. Furthermore, models for a PLOFC accident in two designs (a UO2 and TRISO core) at multiple power levels showed greater temperature margins for the TRISO core at all power levels. Using this information, along with other scoping analyses and constraints, the TCR design team selected a power level of 3 MWth and a TRISO-based core design. For this design RAVEN was applied to vary RIA parameters in RELAP5-3D models at HZP and HFP to understand the effect on figures of merit such as peak power, fuel and coolant temperature, and energy deposition. This sensitivity study found that the inserted reactivity worth was the most important parameter controlling all figures of merit, but for insertion up to 1.5$ no failure of TCR fuel is anticipated. For constant reactivity insertions, the magnitude of the fuel temperature coefficient was found to have the greatest effect on all figures of merit under most circumstances. These results not only demonstrate the anticipated robust safety of the proposed TCR fuel form but also provide a reference for future metal-hydride moderated systems to understand RIA behavior. In the PLOFC, the impact of heat transfer enhancement due to wavy flow channel effects dominated the variance in peak temperatures, and variations in heat exchanger elevation provided the greatest control on natural circulation flow rate. Additionally, no fuel particle failure is anticipated in the PLOFC.
This report presents the design and evaluation of an innovative wall system. This highly insulated (high-R) light-frame wall system for use above grade in residential buildings is referred to as Extended Plate & Beam (EP&B). The EP&B design is the first of its kind to be featured in a new construction test house (NCTH) for the DOE Building America program. The EP&B wall design integrates standard building methods and common building products to construct a high-R wall that minimizes transition risks and costs to builders.
Changes to the LANL / BNL photocathode programs are a strategic response to evolving needs and opportunities within NNSA. The Materials by Design strategy engages the broader material science communities at both LANL and BNL. Progress has been non-linear with time and investment. Techniques honed for one application can be rapidly applied to many others. For example, Cs 2 Te following successes of smooth K 2 CsSb. We enter a cycle of using accelerator-based tools to improve accelerators. Providers of advanced tools of material science (accelerator based x-ray light sources) are now availing themselves of the same capabilities to improve the next-generation machines. We are looking to define future collaborations.
Progress and status reviews allow teams to provide updates and targeted information designed to inform the customer of progress and to help the customer understand current risks and challenges. Both presenters and the customer should have well-calibrated expectations for the level of content and information. However, what needs to be covered in systems-level management reviews can too often be poorly defined. These unclear expectations can lead teams to overpreparing or attempting to guess what information the customer considers as most critical. This aspect of the review process is stressful, disruptive, and bad for morale – and time spent overpreparing reports is time spent not focusing on the technical work necessary to stay on schedule. To define and address these issues, this report was designed to observe various aspects of development program coordination and review activities for NNSA and Navy customers, and then to conduct unbiased, independent Human Factors observation and analysis from an outside perspective. The report concludes with suggestions and recommendations for improving the efficiency of information flow related to reviews, with the goals of increasing productivity and benefitting both Sandia and the customer.
This memo documents the mechanical loading analysis performed on the second set of DTOcean program WP4 foundation and anchor systems submodule design iterations [4]. Finite Element Analysis (FEA) simulations were performed to validate design requirements defined by Python based analytic simulations of the WP4 program Naval Facilities Engineering Command (NAVFAC) tool. This FEA procedure focuses on worst case loading scenarios on shallow gravity foundation and pile anchor designs produced by WP4. These models include a steel casing and steel anchor with soft clay surrounding the steel components respectively.