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DOE OSTI2021

Summary of DASSH-CFD Inter-Assembly Heat Transfer Comparison

Fast reactor core design requires integration of detailed power distributions obtained from neutronics models and thermal fluids calculations to determine temperature distributions and flow conditions. During the design phase, calculations may be repeated as specifications change and computational expense must be balanced with solution accuracy, creating an opportunity for the use of intermediate-fidelity codes. To that end, the Ducted Assembly Steady-State Heat Transfer Software (DASSH) has been developed at Argonne National Laboratory to perform full-core, steady state coolant and fuel pin temperature calculations in reactors consisting of hexagonal, ducted assemblies. DASSH is a first-order, finite volume subchannel code based on methodologies initially introduced in the legacy thermal hydraulics codes SUPERENERGY-2 and SE2-ANL. It can work coupled with the Argonne Reactor Computational (ARC) code suite to obtain neutron and gamma power distributions based on neutron transport solutions and features many improvements and new capabilities compared to its predecessors. DASSH calculates temperatures by balancing energy between subchannels and relies on correlations to account for flow distribution and coolant mixing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Mechanisms Engineering Test Loop (METL) Experimenter's Guide

The Mechanisms Engineering Test Loop (METL) was built to streamline and accelerate the in-sodium testing of systems and components under conditions that simulate a sodium-cooled fast reactor pool environment. The METL team at Argonne National Laboratory (ANL) can assist experimenters in achieving their technical goals by providing liquid-metal expertise and access to infrastructure required for most alkali metal related research. This document offers a brief overview of METL and provides a basic design guide for researchers interested in conducting research at the facility. Additional information regarding the history and operations of METL can be found in §6.1. Furthermore, high resolution images found in this document as well as CAD files of aforementioned vessels can be provided upon request.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Testing of In-Sodium Hydrogen Sensor

The goal of this project is to demonstrate diffusion-type in-sodium hydrogen meters (ISHMs) for real-time, in-situ monitoring of hydrogen concentration in molten sodium of a sodium-cooled fast reactor (SFR). The FY22 R&D efforts have been focused on (a) enhancement and fabrication of meshed ISHM prototype, (b) reconstruction of ISHM sodium test apparatus, (c) in-sodium test of meshed ISHM, and (d) ISHM performance and failure evaluations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

A Very High Temperature Lead Cooled Microreactor Concept using Advanced Materials and Thermophotovoltaics Direct Energy Conversion Technology

Recently, significant advances have been made in designing high-temperature capable nuclear fuels. In parallel, important advances have occurred in design of nanophotonic structures (coatings) with ability to shape the frequencies of light being emitted from hot surfaces. In this project, we bring these materials advances together into a design for a new direct conversion nuclear powered system with high thermodynamic efficiency, inherent safety, and unprecedented reduction in size and weight relative to current designs. The concept involves using the heat generated from nuclear fission to raise the temperature of a selective thermal emitter to >800°C. The selective emitter shifts the normal broad brand emissions of light from its hot surface into the correct near- and mid-IR bands that produce electricity with inexpensive photovoltaic cells. We have designed a new annular flow Lead Fast Reactor (LFR) that minimizes overall reactor/power conversion footprint. The results indicate that the overall microreactor/power-system design offers a 2X reduction in size over existing technology that is based on the sCO 2 reverse compression Brayton cycle. Additionally, the only moving part in our design is a circulation pump to cool the PVT panels.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Gear Test Assembly - Experimental Testing and Analysis of Gears and Bearings - FY2023

The Gear Test Assembly (GTA) is an experimental test apparatus built and installed in the Mechanisms Engineering Test Loop (METL) at Argonne National Laboratory (ANL). The focus of GTA has been to test larger radial spur gears sets machined from Inconel 718 along with a variety of bearing components that are to be used in the fuel handling machines built for use in pool-type sodium fast reactors. The testing in GTA also informs the design and material choices for the full-scale Gripper Test Assembly (GrTA) being developed by Argonne. To date, the Inconel 718 gears used in all six GTA experimental campaigns have completed over 23.2 million revolutions under various loads equivalent to approximately 37,156 simulated fuel assembly maneuvers (removal from and reinsertion to core) using a conventional height of a core assembly. While the heat treated and untreated tapered roller bearings and cylindrical pin thrust bearings used in the first five experimental campaigns have experienced mechanical or material failure after completing a range of fuel assembly maneuvers ranging from as early as 575 operations in Campaign #5 to 9800 operations in Campaign #1, Campaign #6 was the most successful campaign, achieving 22,563 maneuvers and was concluded before any catastrophic failure occurred. In this report we will present the results from the most recent campaigns, #5 and #6. The longevity of Campaign #6 was likely aided by a variety of adjustments made to the materials and components used and to the experimental testing procedures employed. In terms of materials and components, new gearboxes were fabricated, as opposed to using the old gear boxes which had been impacted by the previous bearing failures, and ball bearings were replaced with tapered roller bearings. The adjustments to the experimental procedure focused on a lowering of the maximum torque applied to the gears and bearings and the increase of the duration of each simulated fuel assembly maneuver conducted. This report will present and discuss the changes made to the materials, components, and procedures and the subsequent results for Campaign #5 and Campaign #6.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

A mechanistic model for creep and thermal aging in Alloy 709

This report describes a physics-based model for creep and thermal aging in Alloy 709. Alloy 709 is an advanced austenitic alloy, targeted for use in future Sodium Fast Reactors (SFRs) and other advanced reactors. The material has superior high temperature properties compared to currently qualified 316 and 304 stainless steels. However, the available creep and thermal aging test database for Alloy 709 is significantly more limited compared to the historical materials. The physics-based model developed here is one way to accelerate the qualification of the material by providing more accurate long-term predictions for creep properties and thermal aging, compared to current empirical time-extrapolate techniques. The crystal plasticity finite element model is used to predict the deformation and failure of alloy 709. The same setup for the CPFE model is used in both the baseline model calibration process and the simulation campaigns for parameter inference. Specific constitutive choices are made for Alloy 709 to capture the primary deformation mechanisms. The dislocation creep formulation developed by Hu and Cocks is extended to account for coupled precipitation formation and the grain boundary cavitation model developed by Sham, Needleman, et al. is used to model grain boundary cavitation-induced failure. A novel update algorithm is proposed to render the semi-discrete constitutive update for the Sham-Needleman model unconditionally stable. A progressive calibration approach is adopted based on the observations that several types of material responses can be effectively decoupled. A surrogate model is trained based on full-fledged CPFE simulations to accelerate the forward model evaluations, and stochastic variational inference (SVI) is used to calibrate the unknown microstructural model parameters. The calibrated mechanistic model is used to predict the long-term creep life of Alloy 709, and the predictions are compared against classical empirical approaches.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Progress on Design and Production of Oxide Dispersion–Strengthened Alumina-Forming Austenitic Alloys for Nuclear Applications

Alumina-forming austenitic (AFA) alloys are a promising class of nuclear materials due to their high-temperature oxidation/corrosion resistance and mechanical properties. Unfortunately, these alloys are limited in use for core material applications, specifically in their use as nuclear fuel cladding. Improving these alloys through a fine dispersion of oxide precipitates and thus increasing the effective irradiation sink strength of the alloy system may mitigate many of the degradation phenomena expected during alloy deployment (such as high-temperature helium embrittlement for lead-cooled fast reactor applications). This work package has started using a combination of conventional and advanced manufacturing approaches to the fabrication of oxide dispersion–strengthened (ODS) AFA materials. As such, this report summarizes progress to date in the fabrication of new compositions of these ODS AFA materials with a focus on reactor-specific design and a multifaceted manufacturing approach. Two different compositions of AFA gas-atomized powders were procured, and 200 g of each have been prepared using mechanical alloying with the expectation of high-temperature consolidation in Q1 of FY24. In addition, an approach to use reactive cover gases to promote additional precipitate formation has been developed for implementation in FY24 as a direct comparison with conventionally manufactured alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Gripper Test Assembly Status of Fabrication and Assembly Report – FY2023

The Gripper Test Assembly (GrTA) is an experimental test article designed for use at the Mechanisms Engineering Test Loop (METL) at Argonne National Laboratory (ANL). The GrTA will test mechanical components intended for use in compact in-vessel fuel handling machines (FHM) for sodium cooled fast reactors (SFR). The GrTA will test radial and thrust ball bearings, radial and thrust roller bearings, ball screws and nuts, spline shafts and ball nuts, radial spur gears, and universal joints submerged in liquid sodium at high temperatures (250°C-650°C). These components will be tested in a prototypic, full-scale gripper head and under prototypic SFR fuel handling loads. The design of the GrTA was completed in FY2022 and has progressed into fabrication and initial assembly. All custom manufactured components have been fabricated and delivered to ANL following inspection. All custom gears have been fabricated and delivered to ANL. Bearings, ball screws, and spline shafts have been purchased to allow for initial assembly in air, and sodium compatible versions have either been purchased or are in design at appropriate manufacturers. The drive motor systems that operate the various functions of the GrTA have been purchased and delivered to ANL. Instrumentation and control hardware has been purchased and delivered to ANL as well. All sodium facing components of the GrTA have been assembled and show proper fit. The three main functions of the GrTA are to raise or lower a gripper head under prototypic fuel handling loads, rotate the gripper head while carrying a fuel handling load, and extend or retract gripper jaws that lock a fuel assembly to the gripper head. All three functions have been demonstrated in air, while under no load. The GrTA project will move forward next fiscal year by completing the initial assembly, incorporating the drive system and I&C, and preparing the assembly for introduction to sodium.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Preliminary Development of Heat Transfer Model-Based Control Algorithms of Liquid Sodium Purification System: Advanced Sensors and Instrumentation Advanced Controls

Monitoring the operation of sodium purification system is essential for efficient operation of sodium fast reactors. In this work, a heat transfer model has been developed for monitoring the plugging meter and cold trap systems at the Mechanisms Engineering Test Loop (METL) liquid sodium facility at Argonne National Laboratory. The model of the purification system was developed by treating the respective aspects of the cold trap purification loop and plugging meter diagnostic loop as two separate control volumes using information from the METL piping and instrumentation diagram (P&ID). A model predictive controller was designed using first order differential equations with the specified boundary conditions. The system behavior was studied with a tuned optimized procedure using the internal cold trap temperature and plugging meter outlet temperature as control variables, and the air blower temperature as an independent variable respectively. Results of computer simulations obtained in this study compared favorably with experimental data showing very good reference tracking response with negligible overshoot as both plugging meter and cold trap physical models approach the setpoint.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Mechanisms Engineering Test Loop (METL) Experimenter's Guide - Revision 2

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

System Level Analysis Software Validation with Argonne's THETA Experimental Facility (Final CRADA Report)

Under this CRADA, the Contractor will be generating and collecting sodium-based experimental data with modern instrumentation. The Contractor will work symbiotically with the Participant to generate SAS4A/SASSYS-1 and SAM computational models of the THETA experimental facility for validation of the system-level analysis codes utilized by Oklo, Inc. This effort will be specifically focused on the design basis event space scoped by Chapter 15 of the NRC’s Standard Review Plan (NUREG-0800) with an emphasis on natural circulation and thermal stratification phenomena in a prototypic sodium-cooled fast reactor (SFR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

FY24 Efforts to Revive the ANS-54.8 Liquid Metal Fire Protection in LMR Plants Standard

The U.S. has Designed, constructed, and operated several sodium fast reactors (SFRs). To support these efforts and to document best practices, activities associated with the development of standards and consensus standards were carried out, however the vast majority of those SFR standards are currently in a withdrawn or inactive status.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Gear Test Assembly – Experimental Testing and Analysis of Gears and Bearings – FY2024

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Improvements to Passive Heat Removal Systems in SAS4A/SASSYS-1

Safe operation of a sodium cooled fast reactor (SFR) can be significantly aided by heat removal systems that do not rely on active power, but function using passive means. Passive heat removal from such reactors can employ systems that extract heat directly from sodium coolant by utilizing heat exchangers submerged in the coolant and natural circulation of a metal fluid sodium (or sodium potassium eutectic) to heat exchangers (generally air cooled) externally located. These types of systems are generally called Direct Reactor Auxiliary Cooling Systems (DRACS). Another type of passive heat removal system particularly suitable for pool-type SFRs relies on indirect extraction of heat by utilizing air flow in an annular region surrounding the reactor guard vessel which discharges the heat via the chimney effects to the atmosphere. These systems are generally called Reactor Vessel Auxiliary Cooling Systems (RVACS).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Verification of the ENDF/B-VII.1 Based MC 2 -3 Library Rev.1

The MC 2 -3 code, developed by Argonne National Laboratory under the DOE-NE NEAMS program, is a multigroup cross section generation code for fast reactor applications. Last year, the ENDF/B-VII.0 (E70) MC 2 -3 library, which has been extensively used, verified, and validated over a long period, was intensively reverified and updated to support the commercial grade dedication (CGD) requirement of the TerraPower Natrium project. This year, the ENDF/B-VII.1 (E71) MC 2 -3 library, the preliminary version of which was generated several years ago, was regenerated and rigorously verified to support the Natrium project as well as the completion of verification of the E71 library. The E71 library was verified using the process developed during the verification of the E70 library, including comparisons of cross sections with the NJOY-generated cross sections, comparisons of the resolved resonance cross sections with those using the PEDNF library, and comparison of total cross sections with the sum of partial cross sections. Additional verifications were conducted to ensure that the benchmark problem solutions with the E71 library are reasonable compared to the corresponding Monte Carlo solutions. Furthermore, the E71 gamma library was generated, which includes data for prompt gamma, delayed gamma, and delayed beta as well as neutron and gamma heating. The gamma library was verified at the level of individual isotopes. The EBR-II core solutions from MC 2 -3/ DIF3D and MCNP were compared, demonstrating that those solutions in terms of k-effective and assembly powers were in good agreement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Status Update for the Bearing Test Article Design Development – FY2025

Argonne National Laboratory’s (ANL) Mechanism Engineering Test Loop (METL) team has completed the design of a new Bearing Test Article (BTA) to test mechanical bearings for sodium fast reactor (SFR) mechanism applications. Recognition of the need for further bearing testing arose due to failure of rolling element bearings during testing of components for in-vessel fuel handling machines (FHMs) conducted at ANL in the METL facility. While the initial concept of the BTA arose from a need for bearing testing for FHM applications, the BTA design is configurable to test bearings of various sizes, types, and materials submerged in liquid sodium. Requirements for loading, shaft speed, temperature, and bearing size range were chosen based on requirements for SFR fuel handling. Modular components in the test bearing housings and shaft sleeves allow the design to accommodate a wide range of bearing sizes without modification to the larger assembly. Loading will be applied with screw jacks on the air side of the test article that will be adjusted to apply radial, thrust, or combined loading conditions. Applied force will be measured by load cells in-line with the loading mechanisms. In FY25 we have completed the design of the test article that will be fabricated and undergo testing at the METL facility. Drawing packages for the fabrication of the test article have been drafted and are being reviewed to add part tolerances and we have started ordering off-the-shelf components. The report will summarize the development of the Bearing Test Article design starting with an overview of the entire test article, then a more detailed examination of the bearing test section, loading mechanisms, test article ports and seals, and support structures, including analysis that went into significant design choices.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Nanostructured Alumina Forming Austenitic Alloy (NAFA) Production using Advanced Manufacturing

Alumina-forming austenitic (AFA) alloys are well known for their exceptional corrosion-resistant properties due to the formation of a dense oxide scale beneficial in their application as a nuclear material. However, use of the alloys in core material applications is limited due to the high nickel-transmutation and helium generation rate in service, leading to swelling and reduced lifetimes. Enhancing the sink strength to pin gas bubbles and increasing gas management of the material may mitigate many of the degradation phenomena expected during alloy deployment (such as high-temperature helium embrittlement and cavity swelling for lead-cooled fast-reactor applications). Nanostructured materials are usually produced using a time-consuming and expensive batch process like mechanical alloying to achieve the fine homogenous distribution of nanoprecipitates throughout the matrix material. Integrating the nanoprecipitates in modern additive manufacturing processes has proven difficult, as the required nano-sizes or number densities cannot be achieved simultaneously. Efforts to increase the number density of precipitates by adding more precipitate-forming rare-earth elements led to agglomeration of such elements, while the nano-sized precipitates did not form in sufficient quantity when not enough rare-earth element was use. In this work a novel approach to advanced manufacturing and the fabrication of Nanostructured AFA (NAFA) materials was chosen. A dual precipitate-forming NAFA steel with a chemistry more suitable for advanced reactor applications was produced to create an environmentally resistant, high-sink-strength austenitic alloy for advanced reactor cladding applications.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Using FIPD and OPTD to Benchmark Metallic Fuel Performance

This report serves as an introduction, tutorial, and benchmark specification for out-of-pile tests on metallic fuel. It introduces a new user to the EBR-II legacy fuel performance test program and the fast reactor fuel performance databases built to preserve the records. It then details the information stored in each database and how to find it. A benchmark specification is included for a small set of out-of-pile tests on U-10Zr fuel to function as a tutorial demonstrating how the legacy fuel performance data sets stored in the FIPD and OPTD databases can be used together to benchmark fuel performance models for steady-state and transient performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗