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DOE OSTI2020

Guidelines for IAEA Small Specimen Test Techniques Master Curve Fracture Toughness Testing

Under the auspices of International Atomic Energy Agency (IAEA), a coordinated research project (CRP) entitled “Towards the Standardization of Small Specimen Test Techniques for Fusion Applications” has started since 2017. The overall objective of the project is to provide a set of guidelines for small specimen test techniques (SSTT) based on commonly agreed best practices on main test techniques including tensile, creep, low cycle fatigue, fracture toughness, and fatigue crack growth rate. This will act as the first step of a full standardization of the SSTT. Fusion structural materials, i.e., reduced activation ferritic/martensitic (RAFM) steels, are used for testing. In addition, the project will create a comprehensive mechanical property database of RAFM steels tested by SSTT. For the fracture toughness task of the CRP, three testing methods including Master Curve, local approach for ductile crack growth and cleavage fracture, and ductile approach at room temperature will be evaluated. This report focuses on developing guidelines for the round-robin Master Curve testing based on the ASTM standard E1921-19b “Standard Test Method for Determination of Reference Temperature, T 0 , for Ferritic Steels in the Transition Range” and commonly agreed best practice from researchers at Oak Ridge National Laboratory (ORNL), Centre for Energy, Environment and Technology (CIEMAT), and UK Atomic Energy Authority (UKAEA).

36 MATERIALS SCIENCE↗

DOE OSTI2020

RELAP5-3D Modeling of High Temperature Test Facility (HTTF) Test PG-26

The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a scaled integral effects experiment designed to investigate transient behavior in high-temperature gas-cooled nuclear reactors with prismatic fuel and reflector blocks. Several tests have been completed, and more are still planned to at the HTTF, including depressurized conduction cooldown (DCC) and pressurized conduction cooldown (PCC) transients. This report analyses test PG-26, a progression of the Double Ended Inlet-Outlet Crossover Duct Break transient that is referred to as a DCC. PG-26 has been performed at the HTTF between May 30 and June 30, 2019. Core initial conditions (i.e., before the DCC started) have been met using low power (<100 kW) and two of ten available electric heaters. The DCC transient was initiated during the 50 th hour of the test. The break valves were opened, and hot helium from the core and cold helium from the reactor cavity simulation tank (RCST) started mixing. The gases flowed in a countercurrent fashion, where the top half of the hot duct contained hot helium that flowed in one direction and cold helium that flowed in the other direction in the bottom half of the duct. After the pressure and density reached equilibrium, the event entered a diffusion mode. The onset of a reverse natural circulation was not observed during the DCC period of the test. Version 4.4.2ie of the RELAP5-3D computer code has been used to model the HTTF PG-26 test, and results have been compared to available high-quality measured data. The model used in this study is the quality-controlled HTTF RELAP5-3D model (HTTF base 2018-04-19 QA), originally developed by P. Bayless. The report includes RELAP5-3D results of the “base calculations” as well as some sensitivities to important uncertain model inputs, such as primary helium mass flow rate, core ceramic thermal properties, as well as heat evacuation and loop friction models. Using the base RELAP5-3D model predicts a countercurrent helium flow in the hot duct observed at the beginning of the DCC, but instead of going into a molecular diffusion mode, the model predicts the onset of natural convection. Increasing friction in the core and hot duct prevents the natural convection from happening in some of the simulations. Although some temperatures are well predicted (and even overpredicted), the general tendency is to underpredict the ceramic and helium temperatures and heat removal rates during the DCC, resulting in many of the assessment findings being in minimal or insufficient agreement with the data. It is worth noting that the described discrepancies between measured data and RELAP5-3D predictions are not RELAP5-3D code limitations. More so, they reflect limitations in boundary condition and thermal property knowledge. While the RELAP5-3D calculations of the test provide some insights into what happens during the transient, and point to missing or potentially uncertain data to which the experimenters can direct their attention, the principal conclusion is that the PG-26 test data are insufficient for a system code assessment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Flow Sensor Test Article (F-STAr) - Assembly and Water Testing Report

The Flow Sensor Test Article (F-STAr) is a new test article under development for the Mechanisms Engineering Test Loop (METL) facility at Argonne National Laboratory. F-STAr’s purpose is to provide sodium submersible, high flowrate testing capabilities for the development of sensors, components, fluid studies and more. This report will provide a status update on the assembly and initial qualification testing of F-STAr. All components have been manufactured and received. Some modifications to these components were made to support minor changes in the test article design. Additionally, the power distribution, and data acquisition and control enclosures were designed, completed, and tested along with the control system software. Furthermore, initial qualification testing was completed in water. Finally, the F-STAr submersible Electromagnetic Flowmeter (EMFM) was completed and calibrated in flowing sodium.

42 ENGINEERING↗

DOE OSTI2024

Test Stand Functional Requirements for Testing MQXFA Magnets and Q1/Q3 Cryostat Assemblies (Rev.3)

This document specifies test stand requirements for testing the High Luminosity LHC (HL-LHC, or HiLumi LHC) MQXFA magnets and LQXFA/B cryostat assemblies. Total of 12 (3 pre-series, 7 series production and 2 re-work) of these cryostat assemblies are expected to be fabricated and delivered to CERN by the U.S. HL-LHC Accelerator Upgrade Project (US HL-LHC AUP) as part of the U.S. contributions to the LHC High Luminosity Upgrade. These cryostat assemblies are the quadrupole magnetic components of the HL-LHC Q1 (LQXFA) and Q3 (LQXFB) inner triplet optical elements in front of the interactions points 1 (ATLAS) and 5 (CMS). One LMQXFA cold mass is installed in each cryostat assembly and each cold mass consists of two MQXFA quadrupole magnets. MQXFA, LMQXFA and LQXFA/B functional requirements are specified in [1,2,3].MQXFA magnets will be tested at BNL Vertical Magnet Test Facility (VMTF) and 12 cryostat assembly tests planned at Fermilab’s Horizontal Magnet Test Facility (HMTF). This document specifies the test stand functional requirements for testing MQXFA magnets and LQXFA/B cryostat assemblies.

43 PARTICLE ACCELERATORS↗

DOE OSTI2025

Preliminary Plan to Inform Testing of a Heat Exchanger Test Article

This report presents a preliminary plan to guide the qualification testing of advanced heat exchanger (HX) components for nuclear-to-industrial heat transfer applications. The objective is to establish a defensible, physics-based methodology that integrates computational modeling, targeted experimentation, and in-service inspection considerations to demonstrate component performance and reliability under representative reactor conditions. The analysis identifies Sodium-cooled Fast Reactor (SFR) and High-Temperature Gas-cooled Reactor (HTGR) systems as reference configurations in terms of temperature, pressure, and chemical environment. Within these operating envelopes, dominant degradation mechanisms— including creep–fatigue interaction, flow-induced vibration, corrosion, and diffusion-bond deterioration—were evaluated to define test requirements. A comprehensive computationalexperimental framework is proposed to support life prediction and qualification activities. The framework couples high-fidelity structural-mechanics, thermal-hydraulic, and fluid-structure interaction models with accelerated degradation testing to produce a traceable linkage between microstructural evolution, mechanical performance, and remaining useful life (RUL). The approach adheres to established Verification, Validation, and Uncertainty Quantification (VVUQ) standards (ASME V&V 10/20; NUREG-2152) and incorporates a digital-twin architecture for continuous model refinement through data assimilation. The plan further outlines testing methodologies, including pre-test analyses, test-loop design parameters, and sensor placement strategies that maximize information yield while maintaining mechanistic fidelity. Complementary sections describe in-service inspection (ISI), on-line monitoring (OLM), and structural-health-monitoring (SHM) techniques applicable to compact HX geometries typical of advanced reactors. Collectively, these activities establish the technical foundation for demonstrating 40-60-year equivalent service life of advanced heat exchangers in support of the U.S. Department of Energy’s Advanced Reactor and Integrated Energy Systems programs. The forthcoming phase will execute the defined pre-test analyses, initiate hardware fabrication, and implement the integrated testing campaign to validate the proposed qualification methodology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Testing of a 40-kWth Counterflow Particle-Supercritical Carbon Dioxide Narrow-Channel, Fluidized Bed Heat Exchanger

Particle-based primary heat exchangers (HXs) must deliver sCO2 fluid temperatures above 700°C to couple particle-based concentrating solar receivers and thermal energy storage (TES) sub-systems with efficient sCO2 power cycles. Particle-sCO2 HX designs have struggled to meet DOE cost targets (≤ $150/kWth) due to the amount of expensive nickel alloys necessary for manufacturing full-scale, particle-sCO2 HXs. Our team has demonstrated that mild bubbling fluidization of falling particles in a counterflow narrow-channel fluidized bed can reduce required HX surface area and thus, costs by increasing particle-wall heat transfer coefficients hT,w > 800 W m-2 K-1. This paper reports on the fabrication and testing of a stainless steel, particle-sCO2 HX with 12 fluidized-bed channels approximately 10.5 mm deep spaced between diffusion-bonded, micro-channel sCO2 plates. The HX with a core length of ≈0.56 m is fed with CARBOBEAD HSP particles through a short, fluidized freeboard zone just above the core. Testing to date in the National Solar Thermal Test Facility (NSTTF) at Sandia National Laboratories has shown that parallel bed fluidization maintains uniform particle inventory across the instrumented channels. Heat transfer thermal duty between the particle and sCO2 flows exceeds 30 kWth with sCO2 inlet temperatures of 200ºC and particle inlet temperatures up to 440ºC and mass flow rates of 0.2 kg s-1 fluidized by counterflowing gas flow rates of 0.005 kg s-1. Tests at higher particle and sCO2 inlet temperatures (600ºC and 400ºC respectively) are targeted to achieve > 40 kWth with model-predicted overall heat transfer coefficients U > 400 W m-2 K-1.

14 SOLAR ENERGY↗

DOE OSTI2026

Thermal Performance of Spandrel Assemblies in Glazed Wall Systems: Laboratory Test Design – Challenges and Test Results

Accurate thermal performance calculation procedures for opaque spandrel areas in curtain wall and window wall systems are essential for rating systems when comparing spandrel systems. However, there is a lack of consensus in thermal modeling needed for accurately characterizing heat transfer through spandrel assemblies due to the complex arrangement of materials and structural components. Several studies indicate that conventional 2D thermal simulations may overestimate R-values by 30% compared to physical testing and 3D simulations. Detailed simulations and well-curated laboratory test data are necessary to build confidence in simulation models, which will later be used to develop correlations to improve widely used conventional 2D thermal simulations. This study aims to experimentally test heat transfer through various spandrel assemblies to validate 3D simulation models. Also, the challenges of conducting a thorough testing design along with the solutions would be documented. The team developed a design for testing spandrel assemblies, making appropriate modifications to the existing heat, air, and moisture (HAM) chamber to accommodate the testing needs. Two moveable baffles were designed and fabricated to guide airflow direction parallel to the test article surface. The data acquisition capabilities in the chamber were upgraded to add more than two hundred sensors to the climate and indoor side of the chamber. The goal is to provide a quality dataset for validating complex 3D modeling simulations, which will be used to develop improved thermal simulation techniques that more accurately represent the thermal behavior of spandrel assemblies and their integration within the building envelope. This paper will summarize the results for the boundary conditions of the testing and the temperature variation across different locations of the spandrel assemblies.

Kunwar, Niraj [ORNL] (ORCID:0000000263457652)↗

DOE OSTI2020

Development of a Fast Spectrum Test Loop at the Advanced Test Reactor

The United States currently has no infrastructure to provide a fast neutron flux for research purposes. With the increased interest in Sodium Fast Reactor (SFR) technology, this type of infrastructure is necessary for further development of key components such as Versatile Test Reactor. The purpose of this research is to characterize the thermal hydraulics of a test loop capable of providing SFR like operating conditions. This includes showcasing feasibility in mechanical design, reaching a high fast to thermal neutron flux ratio ideally near 15, an inlet temperature of 600 K, a temperature increase of 50 K from inlet to outlet, and test chamber radial temperature profiles similar to SFR fuel rods. Existing infrastructure must be used to provide the neutron flux for the test loop, and it was decided that the Advanced Test Reactor (ATR) was the optimal choice due to its availability and previous work conducted similar to this project. To achieve the goals of this project, several mechanical designs were put forward with the most feasible design being chosen. The neutronics and thermal hydraulics analysis were conducted in tandem, with an iterative process used to maximize efficiency of both systems. Analysis of the neutronics system was done using Attila and MCNP while the thermal hydraulic analysis was done using RELAP5. This document outlines the thermal hydraulic analysis, though it includes the results from the mechanical design and neutronics analysis. The loop design includes a primary system with molten sodium, a set of “booster fuel plates” to increase the fast flux within the test section, a water system to cool the booster fuel plates, a helium system to separate the sodium and water systems, and a thermal neutron shield to decrease the thermal flux. Each of these systems is modeled in the neutronics and thermal hydraulic analyses. The RELAP5 model showcased the feasibility of the design proposed by this report. For all designs tested, the desired 50 K temperature increase from inlet to outlet was reached. The fuel rod radial temperature profile was lower than was expected, likely due to a lower rod thermal power than expected. A parametric study was also conducted to characterize the system with three fuel pins. This study showed the effect of thermal power, fluid pressure, fluid velocity, and inlet temperature on the system. ©Copyright by Anton Higgins December 11, 2019 All Rights Reserved

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2024

Evaluation of Additively Manufactured Monolithic SiC and SiC-Matrix Ceramic Matrix Composites for Concentrating Solar Receiver Applications - Fabrication and Testing of Receiver Design Feature Specimens in a Simulating Lab Test via Laser Heating

Increasing operating temperatures of solar receivers is paramount to the efficiency of concentrated solar thermal (CST) and solar power (CSP) systems. Owing to its high temperature stability combined with excellent thermal and optical properties, SiC has been the material of choice for application in high-temperature solar receivers. We report the results of our study of the effective heat transfer characteristics of several candidate SiC structure motifs, or feature geometries, which are fabricated via additive manufacturing. The SiC structure motifs studied include different permutations of three-dimensional periodic lattices and defined shapes. A solar-thermal simulating laboratory test setup is constructed using a 4kW CO2 laser system with beam shaping optics to apply concurrent radiative heating power on one face of 2”-diameter cylindrical feature specimens, representing the structure motifs of interest for receiver element design, while flowing through the sample as heat transfer fluid. Using the test setup, simulative test conditions representative of a concentrated solar flux of up to ~2000 suns could be achieved in the lab tests under varying air flow through the test structure. A simple 1D numerical analysis scheme is developed to extract an effective or compound heat transfer coefficient representative of the test structure under steady-state heat flow conditions. The test results and their use to guide the selection and optimization of SiC material and structure motifs for the receiver element design fabrication are discussed.

14 SOLAR ENERGY↗

DOE OSTI2024

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly

DOE manages over 300 types of spent nuclear fuel (SNF), many of which are located at the Idaho National Laboratory (INL) site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in onsite storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale road-ready dry storage program at the INL site. In support of establishing a large-scale road-ready dry storage program at the INL site, the Road-Ready Demonstration will first package Fort St. Vrain SNF currently in dry storage at INL into several DOE Standard Canisters (DOESCs). These DOESCs will in turn be loaded into another containment similar to commercial multi-purpose canisters. This multi-purpose canister will then be compatible with a transportation or storage system, such as a storage cask for interim storage or transportation package for offsite transport. These DOESCs will remain sealed over the course of their storage, transportation, and applicable disposal functions. The closure process for the DOESC will include fuel and basket loading, welding, inspection, leak testing, and, if needed, repair. As a follow-up to previous discussions on the design of the DOE Closure Leak Test Assembly (LTA), this report describes recent fabrication and testing efforts performed at INL. DOESCs are sealed by two sequential gas tungsten arc welds, both of which are performed by remotely operated and semiautomatic welding systems. The first weld is a circumferential pipe weld that completes the assembly of the canister body and lid assembly. The second and final closure weld attaches the vent plug to the vent socket via a butt joint. After the second weld is performed, the welds are helium leak tested using an evacuated envelope technique. The LTA was designed for both remote and manual operation. This report describes the fabrication and performance testing associated with the evacuated envelope technique. INL staff designed, fabricated, and tested the LTA at INL facilities. This testing included establishing technique and system sensitivities in accordance with ASME and American National Standards Institute N14.5 requirements. Forthcoming work will cover such areas as design optimization, process and personnel qualification, and implementation in Road-Ready Demonstration operations.

42 ENGINEERING↗

DOE OSTI2024

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly

DOE manages over 300 types of SNF, most of which are located at the INL site. “Road-ready dry storage” is a management concept where SNF is packaged into dry, sealed canisters, which are then placed in on-site storage in anticipation of later removal. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE SNF for road-ready dry storage. In support of establishing a large-scale road-ready dry storage program at the INL site, the demonstration will package a select amount of DOE-managed SNF into DOE Standard Canisters. The closure process for the DOE Standard Canisters will include fuel and basket loading, welding, inspection, leak testing, and if needed, repair. As a follow-up to previous discussion on the design of the DOE Closure Leak Test Assembly, this report describes recent fabrication and testing efforts performed at INL. DOE Standard Canisters are sealed by two sequential gas tungsten arc welds. Both are performed by remotely operated and semi-autonomous welding systems. The first weld is a circumferential pipe weld that completes assembly of the canister body and lid assembly. The second and final closure weld connects the vent plug to the vent port with an identical butt joint to the circumferential pipe weld. After the second weld is performed on the vent port, these welds are helium leak tested using an inside-out technique. In addition to the commercially available vacuum and leak detector systems, the DOE Standard Canister Closure Leak Test Assembly was designed for both remote and manual operation. This report describes fabrication and performance testing associated with the inside-out technique. INL staff designed and tested systems to accomplish these tasks. Hardware fabrication occurred at INL facilities. Forthcoming work includes design optimization, integration to existing systems, and implementation to packaging demonstration operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly – 24139

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Demonstration of the surveillance test article for an advanced reactor surveillance test program

Operational environments in generation IV reactors involve corrosive and irradiative conditions at elevated temperatures. Typical reactor operations consist of transients which impose cyclic loads on reactor components. These cyclic loads, combined with corrosive and irradiative environments, result in synergistic degradation of component materials. However, limited data exists on the coupled damage effects on materials for reactor environments. While the surrogate material surveillance concept has been used in light water reactors to assess irradiation damage, existing material surveillance technologies are not suitable for in-situ monitoring of coupled material degradation. The materials surveillance program focuses on material degradation management and the estimation of remaining life of reactor components through surveillance test articles. This paper presents the design and analysis methodology of a bi-metal surveillance test article, which uses difference in thermal expansion coefficient between two metals to induce in-situ cyclic loads. This report presents the work conducted in FY 25, to test the surveillance test article in air and salt environments. These test specimens were evaluated after thermal cycle exposure and remaining life is measured through creep test.

36 - MATERIALS SCIENCE↗

DOE OSTI2025

Development of a Miniature Small Punch Creep Test for Post Surveillance Test Specimens

Summary: This poster outlines a new small punch creep test (SPCT) for examining the mechanical properties of post surveillance test specimens. A surveillance test article was previously designed to study the effects of Molten-Salt conditions on the structural materials in the reactor. To better understand the post-test mechanical properties of the specimen, a miniature SPCT is proposed. The principle, testing standards and challenges of the method is introduced in the poster. An initial design of the testing apparatus is shown, and an experiment plan is outlined.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Mechanisms Engineering Test Loop - Flexicask Design and Testing Report

This report documents the design development and testing of the Flexicask system for the Mechanisms Engineering Test Loop (METL). Currently, METL has completed Phase I construction and has been operating for over a year. The Flexicask system (Figure 1) was designed to allow for the removal and insertion of test articles into a METL test vessel while maintaining an acceptable inert atmosphere to minimize contamination of the METL sodium. This report discusses the design development of the Flexicask system, the commissioning and testing of the system, and the initial use of the Flexicask system to remove the Gear Test Assembly (GTA) from the METL Test Vessel 1.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Mechanisms Engineering Test Loop (METL) Operations and Testing Report (FY2021)

This report documents the operations and testing that was performed at the Mechanisms Engineering Test Loop (METL) during FY2021. The METL facility had a very successful third year of operations having logged about 944 days of operations with molten sodium either in a flowing or static condition. Operations were paused in April 20, 2021 to accommodate the refurbishing of building 308’s alkali metal passivation booth and scrubber (AMPB&S). METL is being prepared to support a second larger experiment, the Thermal Hydraulic Experimental Test Article (THETA) and expects to resume the Gear Test Assembly (GTA) testing. The technology development team is also developing two additional experiments, a gripper test article and a flow sensor test article that are expected to go into METL in late FY22 or early FY23.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Report on Thermal Cycling Testing and Bimaterial Weld Development for a Passively Actuated Materials Surveillance Test Article

This report describes the continuation and conclusion of a thermal cycling test program designed to evaluate a family of passively actuated material surveillance test articles. Additionally, the report details a preliminary study of bimetallic welding between 316 stainless steel and TZM (tungsten-zirconium-molybdenum) alloy. The test program was ultimately successful in demonstrating the basic viability of the passively actuated test article concept as well as specifically validating the procedures developed to size specimens to mimic the mechanical response of operating components. The report discusses and summarizes the results of the complete test program which tested samples for more than 1,000 controlled thermal cycles. The weld development program included both stir friction and electron beam joining. The stir friction welding was ultimately successful, while the electron beam welding would require additional development. Ultimately, these test programs pave the way for implementing a materials surveillance program in future molten salt reactors (MSRs). The report conclusions outline a plan for transforming the research and developed described here into a definite set of practices and procedures for a surveillance program at a future operating reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Test Report for DPP-1 Regulatory Compliance Testing. Volume 1 - Main Report

Consolidated Nuclear Security, LLC (CNS), the management and operating contractor for the Y-12 National Security Complex (Y-12) for the National Nuclear Security Administration (NNSA), is currently responsible for the regulatory testing of a new Type B fissile material shipping package called the Defense Programs Package (DPP)-1. The DPP-1 is a drum packaging with an inner liner and a removable lid. Both the drum body and the lid are filled with an impact-limiting, thermal-insulating material that protects the inner containment vessel (CV) during normal conditions of transport (NCT) and hypothetical accident conditions (HAC) as defined by 10 US code of Federal Regulations (CFR) 71. In fiscal year 2021, the Oak Ridge National Laboratory (ORNL) Package Testing Program (PTP), located at the National Transportation Research Center (NTRC), conducted Type B performance tests as prescribed in Title 10 Code of Federal Regulations Part 71 (10 CFR 71) on seven DPP-1 prototype units. This test report documents test unit (TU) preparation, pretest condition, conditioning of each TU, NCT and HAC testing, and posttest measurements and observations made of the damage resulting from these tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗