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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
NASA NTRS1977

Development of flame resistant treatment for Nomex fibrous structures

Flame resistant fibrous materials for space shuttle application were developed through chemical modification of commercially available aromatic polyamide fibrous products. The new surface treatment was achieved in the laboratory by ultraviolet activation of the fabric in the presence of fluoroolefin monomers and a diluent gas. The monomers grafted under these conditions provide the improved properties of the fabric in flame resistance, chemical inertness, and nonwettability without the sacrifice of color or physical properties. The laboratory reaction vessel was scaled-up to a batch continuous process, which treats ten yards of the commercial width textiles. The treated commercial width Nomex (HT-10-41) from the scaled-up reactor is self-extinguishing in an oxygen-enriched environment, water-repellent, soft, silky, and improved in chemical resistance. Unlike most textile processes, the grafting unit operates under dry conditions and no chemical by-products have to be washed out of the finished product.

Toy, M. S.↗
DOE OSTI2024

Expansion of the Fuel Motion Monitoring System

The Idaho National Laboratory (INL) Transient Reactor Test Facility (TREAT) Fuel Motion Monitoring System (FMMS) is a diagnostic tool used to visualize the movement and location of fuel within the nuclear reactor during transient experiments. TREAT is designed to test the behavior of nuclear fuel under accident conditions, such as a rapid power increase. The TREAT FMMS, also referred to as the “hodoscope” because of the system’s massive steel collimator component, is located on the North beam port of the reactor. While the FMMS system is capable of incorporating a total of 360 sensors, initial refurbishment efforts resulted in a array of 96 fast neutron detectors. This initial array was configured to provide a narrow (2 channels) full-length view of experiment vessels with an extended 4-channel–wide region in the center of the array capable of providing a full view of all initially planned experiments. Planning and efforts to expand the FMMS array began immediately following the restart of TREAT. An additional 96 detectors were evaluated, characterized, and prepared for installation at TREAT during the summer of 2018. Unfortunately, funding reductions forced the project towards a 5-year hiatus. In October of 2023 funding from the Department of Energy’s Nuclear Science User Facilities enabled the expansion to recommence. This report serves to document the activities performed to expand the FMMS detector array to 192 channels in order to provide a broader view of the larger and more sophisticated test capsules currently planned for future irradiation in the TREAT reactor. This expansion also included doubling the data acquisition capability, addressing the new heat load on the system, and synchronizing the time for all digital components.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Fusion Power measurement at ITER

Nuclear fusion research aims to provide energy for the future in a sustainable way and the ITER project scope is to demonstrate the feasibility of nuclear fusion energy. ITER is a nuclear experimental reactor based on a large scale fusion plasma (tokamak type) device generating Deuterium - Tritium (DT) fusion reactions with emission of 14 MeV neutrons producing up to 700 MW fusion power. The measurement of fusion power, i.e. total neutron emissivity, will play an important role for achieving ITER goals, in particular the fusion gain factor Q related to the reactor performance. Particular attention is given also to the development of the neutron calibration strategy whose main scope is to achieve the required accuracy of 10% for the measurement of fusion power. Neutron Flux Monitors located in diagnostic ports and inside the vacuum vessel will measure ITER total neutron emissivity, expected to range from 1014 n/s in Deuterium - Deuterium (DD) plasmas up to almost 10{sup 21} n/s in DT plasmas. The neutron detection systems as well all other ITER diagnostics have to withstand high nuclear radiation and electromagnetic fields as well ultrahigh vacuum and thermal loads. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

The control of the deuterium–tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10–200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1–4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Material Plasma Exposure eXperiment (MPEX) High Heat Flux Bellows for Component Alignment in High Microwave Environment

The Material Plasma Exposure eXperiment (MPEX) device is a linear plasma device developed to perform plasma material interaction experiments under the conditions prototypic of a fusion reactor divertor. MPEX has multiple systems that must be precisely aligned to the plasma axis, including an electron cyclotron heating system that emits up to 400 kW of microwave power into the vacuum vessel. Five distinct systems require precise alignment on the MPEX device, thus requiring four bellows, all of which are adjacent to the plasma at a relatively high heat flux of approximately 47 kW/m2 and microwave power regions. The MPEX high heat flux bellows (HHFB) is designed to deliver 6 degrees of freedom positioning. The HHFB includes titanium–zirconium–molybdenum (TZM) inserts that are brazed into a Glidcop AL-15 body using a high-temperature braze alloy, thus blocking direct line of sight to an edge-welded bellows from the plasma and microwave screen to block microwaves from the bellows. Further, a custom ConFlat knife edge is machined into the Glidcop AL-15, so the vacuum flanges do not need a braze or weld joint on the vacuum interface. Fingerstock or copper mesh is used to restrict microwave power from entering the interstitial space between the water-cooled Glidcop AL-15 body and edge-welded bellows. Glidcop AL-15 was selected as the material of choice for the water-cooled body because it can maintain mechanical integrity at elevated temperatures, and it also allows for a high-temperature braze. TZM was selected for its machinability and compatibility with vacuum and plasma requirements. A test article is also being considered for a similarly shaped component, the MPEX limiter, that will demonstrate the integrity of the braze joint under high thermal load. Results from this testing will be extrapolated to deduce the lifetime and integrity of the HHFB design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Development of Grade 91 inelastic model for incorporation in ASME Division 5

This report describes a unified viscoplastic model for Grade 91 suitable for use with the Section III, Division 5 design by inelastic analysis procedures in the ASME Boiler and Pressure Vessel Code. The report identifies several key features a successful design inelastic model must capture along with critical aspects of the response of Grade 91 for elevated temperature loading conditions typical of expected reactor operating conditions. An inelastic model for Code design must capture the average response of a material including heat-to-heat and product form variations. This work develops a method for calibrating such a model to a large experimental database using genetic algorithm optimization. Finally, the report compares the final model to several specialized experimental tests to validate that it will be suitable for use in high temperature structural design. The techniques developed here can be applied to the remaining Section III, Division 5, Class A materials to develop a nonmandatory appendix to the Code describing inelastic material models suitable for use with the design by inelastic analysis procedures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Test Plan for the Horizontal Dry Cask Simulator

The thermal performance of commercial spent nuclear fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission(NRC). Canistered dry storage cask systems rely on ventilation between the inner canister and the overpack to convect heat away from the canister to the surrounding environment for both horizontal and vertical configurations. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of the investigation described in this test plan is to produce data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator(HDCS) has been designed to generate this benchmark data and add to the existing knowledgebase. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure(MAWP)rating of 2,400 kPa at400 °C. An existing electrically heated but otherwise prototypic boiling water reactor(BWR), Incoloy-clad test assembly will be deployed inside of a representative storage basket and canister. An insulated sheet metal enclosure will be used to mimic the thermal properties of the concrete vault enclosure used in a modern horizontal storage system. Radial and axial temperature profiles along with induced cooling air flow will be measured for a wide range of decay powers and representative(and higher)cask pressures using various backfills of helium, argon, or air. The single assembly geometry with well-controlled boundary conditions simplifies computational requirements while preserving relevant physics. The proposed test apparatus integrates all the underlying thermal-hydraulics important to defining the performance of a modern horizontal storage system. These include combined-mode heat transfer from the electrically-heated assembly to the canister walls and the primarily natural-convective heat transfer from the canister to the cooling air flow passing through the horizontal vault enclosure. The objective of the HDCS is not to reproduce the performance of a commercial dry storage system for any given set of operational parameters. Rather ,the objective is to capture the dominant physics in a well-characterized test apparatus. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate is of particular importance. While incorporating the best available information based on thermal-hydraulic scaling arguments as well as previous vertical testing, this test plan is subject to changes due to improved understanding or from as built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Deuterium Retention in Neutron-Irradiated Single-Crystal Tungsten

A critical challenge for burning plasma operation is to predict tritium behavior (retention and permeation) in plasma-facing components under intense high heat, particle and neutron flux conditions to minimize the in-vessel tritium inventory and the annual operational tritium release to the public and environment. The Task 3 of the US-Japan Joint Project on Technological Assessment of Plasma Facing Components for DEMO Reactors (PHENIX) aims at advancing the scientific understanding of tritium retention in neutron-irradiated tungsten and tungsten alloys for fusion fuel cycle and safety. A combination of thermal desorption spectroscopy, nuclear reaction analysis, and rate-diffusion modeling code (e.g. Tritium Migration Analysis Program) are used to elucidate deuterium behavior in neutron-irradiated tungsten. This paper will discuss one-of-a-kind neutron irradiation under the PHENIX project and recent progress in neutron-irradiated single-crystal and polycrystalline tungsten material irradiated by low-dose neutron and exposed to high-flux deuterium plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

NSTXU Diagnostic Disruption Dynamic Loading Represented by Response Spectra

This article presents the results of transient dynamic simulations of loads due to disruption eddy currents on the NSTXU vacuum vessel. Dynamic loading at diagnostic mounting locations is expressed as response spectra derived from the time history results of the dynamic structural simulations of a variety of disruption scenarios. The disruption simulations draw on a history of the project assessments of worst case disruptions for specific components. Major efforts to assess disruption loading have included the vacuum vessel which is the major structural support for the machine, as well as the passive plates (PPs), high harmonic fast wave (HHFW) antenna, and centerstack casing. Each one of these efforts included transient electromagnetic simulations producing time-dependent eddy current Lorentz loads (and in some cases halo loads) which then were applied to time-dependent structural dynamic analyses intended to obtain the proper dynamic amplification factors. In some instances, the EM model and structural model were identical allowing direct transfer of EM forces to the structural model. In other cases, the EM and structural model were not identical and the vector potential (VP) transfer method was used. The results files from these analyses were available (or re-run) to post process in ANSYS Classic time history postprocessor. In conclusion, the ANSYS command is used to create response spectra from time history data at desired points on the vessel.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Molten Salt Reactor Experiment (MSRE) Lay-Up - 19187

UCOR, an AECOM-led partnership with Jacobs, manages the cleanup of the 2,200-acre East Tennessee Technology Park for its client, the U.S. Department of Energy (DOE), in Oak Ridge, Tennessee. The site was contaminated with radioactive, hazardous and industrial wastes generated by more than 40 years of national defense and energy missions. 2019 is the 50. anniversary of the shutdown of the Molten Salt Reactor Experiment (MSRE). MSRE ran a brief four years in the 1960's but earned an enduring legacy as an innovative reactor technology concept which has enjoyed a renewal of global interest in the past several years. In 2018, it was given elevated status among advanced reactor programs within DoE's Nuclear Energy office. The MSRE was a graphite-moderated, liquid fueled reactor that operated from June 1965 through December 1969. Although it was shut down nearly 50 years ago, certain systems within the reactor have continued to operate to maintain the facility in a safe and stable condition until it can be demolished as part of the comprehensive cleanup of the Oak Ridge Reservation. These systems are composed of a complex network of pipes, valves, gauges and associated equipment that control air filtration and pressure buildup resulting from fluorine gas generated through radiolysis of the fluoride salt which once fueled the reactor. In 2016, an MSRE Engineering Evaluation Team was chartered to assess current facility conditions and develop recommendations to maintain the facility in a safe state until deactivation and demolition commences. Plans are now underway to 'lay up' the MSRE facility until deactivation and demolition can occur. The proposed MSRE Facility Layup will minimize maintenance costs, reduce personnel risk of injury and exposure, provide reliable electric service to the facility and require zero personnel occupancy at MSRE. In the proposed layup approach a new source of electrical power to required loads will be routed through the facility, and new gas processing and ventilation systems will be installed that will be optimized for safe processing of the constituent hazardous gasses that exist in the legacy vessels at MSRE. (authors)

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

Optimization of laser-induced breakdown spectroscopy for liquid sodium analysis

Next-generation nuclear power reactors based on liquid sodium coolant are in development to improve efficiency and safety. Fast breeder reactors could potentially achieve considerably higher fuel burn-up while producing less nuclear waste compared with current commercial technology. Monitoring of impurities in these types of reactors is important in order to prevent blockage of the cooling system or hazardous reactions due to the high reactivity of sodium with water. In this work, laser-induced breakdown spectroscopy is performed on a liquid sodium sample using a custom-built immersion probe system. The probe provides direct line of sight to the sample vessel in argon atmosphere for laser delivery and signal collection. Signatures of sodium, and impurities such as potassium and oxygen have been obtained. The probe height and immersion have a strong influence on the signal level and stability. Argon flow rates of 2 to 5 liters per minute produce no substantial change in signal, suggesting a steady laminar flowing regime. A high signal-to-noise ratio for impurity elements confirms the suitability of this technique for reliable impurity detection and monitoring in the reactor environment.

Gardette, V. [University of Michigan, Ann Arbor, M↗
DOE OSTI2018

CHI Research. Final Technical Report

The spherical tokamak is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. It is highly desirable for the FNSF to be compact in size and simpler in configuration. Since the central solenoid cannot be used for driving current during the steady-state phase, the tokamak design would also be simplified if it could be built without the solenoid, and it would allow access to lower aspect ratio configurations. Great progress has been made in developing Transient CHI as a viable solenoid-free plasma start-up method. These are: (a) The generation of closed flux current in HIT-II and NSTX, and the successful coupling of this current to induction from the central solenoid, (b) the generation of 0.1 MA closed flux current in HIT-II and demonstration of CHI plasma quality similar to that produced inductively, (c) the generation of a record 0.2 MA of closed flux current in NSTX and the coupling of the CHI-produced discharge in NSTX to induction with subsequent transitioning to a high-performance H-mode discharge when neutral beams were injected and (e) CHI-started discharges to which induction has been applied have ramped-up to 1 MA in NSTX with record low flux consumption. Modeling of the CHI process using TSC and NIMROD has led to a much-improved understanding of the physics of magnetic reconnection leading to large volume flux closure. The NIMROD results indicate that this reconnection mechanism scales to arbitrarily high toroidal field and can be either axisymmetric Sweet-Parker like reconnection or it can be enhanced through plasmoid formation. Motivated by the exciting experimental and modeling results from HIT-II and NSTX, we explored the application of CHI to a reactor configuration, and identified methods for enabling CHI capability in a FNSF-sized experiment. In the first, proven, concept referred to as "NSTX-like", the entire blanket structure is insulated from the vessel using insulating supports. In the second configuration referred to as "DIII-D - like", a toroidal insulated electrode is placed on top of the blanket structure to shield the insulator from neutrons, allowing more than six years of full power operations before insulator replacement. The second concept requires an experimental test for concept validation. To test this second concept, during the period of 2013 to 2016, we designed and implemented a new simpler design CHI configuration on QUEST. Initial TSC modeling and experimental results suggests that work on QUEST may lead to a simpler Transient CHI configuration for a ST-FNSF. Transient CHI is now also in the process of being implemented on the PEGASUS ST at the University of Washington to study very high-current start-up potential of transient CHI.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Analyses Supporting Design Review of TREAT Multi-SERTTA Experiment Test Vehicle

The Transient Test Reactor (TREAT) Facility was designed to evaluate nuclear fuels and structural materials under simulated nuclear excursion and transient power/cooling mismatch situations that could be encountered in nuclear reactors [1]. It was historically operated from 1959 through 1994; ongoing activities at Idaho National Laboratory (INL) include resumption of transient testing capabilities to generate experimental data suitable to validate advanced modeling and simulation of these transient conditions, supporting current and next generation power reactor design and nuclear safety [2]. The resumption of transient testing requires new experiment vehicle designs to satisfy testing needs. The most developed experiment test vehicle candidate in TREAT’s arsenal will be the Multi-SERTTA (Static Environment Rodlet Transient Test Apparatuses) [3], which accommodates up to four concurrent rodlets under separate environmental conditions. An overview of the Multi-SERTTA, which would be placed at the center of the TREAT core during use, is shown in Figure 1. In the current design, all four rodlet test vessels are identical except for flux shaping collar material content to meet design needs. As the date rapidly approaches for TREAT restart, there is a need to have the Multi-SERTTA design approved and tested. Test experiments are an integral component of TREAT operations, and will be integrated among initial restart testing and supporting core reactor physics measurements. An integral step in finalization of any test vehicle design for TREAT is the hosting and resolution of a formal design review.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Solving Operational Challenges with INL's Integrated Waste Treatment Unit (IWTU) - 19003

In 2005, the THOR steam reforming fluidized-bed reactor technology was selected by the U.S. Department of Energy for treatment of sodium-bearing waste at the INL. The initial commissioning was typical, until June of 2012 when an over-pressurization event resulted in a significant release of solids. Recovery and subsequent testing systematically dealt with challenges associated with operation of three primary vessels: the denitration and mineralization reformer, the process gas filter, and the carbon reduction reformer. Additional challenges with the following ancillary systems surfaced during testing: the auger/grinder, sample system, and the approach to decontamination of the process vessels prior to outages. The major operational challenges covered fluidization issues, premature refractory failure, high-temperature gaseous filter element blinding, cementitious material formation in solids handling equipment, sampling reliability, and vessel decontamination. The purpose of this paper is to share the lessons learned during the extended commissioning of the Integrated Waste Treatment Unit. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Filter 'Cook-Off': Filter Selection via Side-by-Side Testing at Actual Process Conditions - 20040

In 2005, the THOR steam reforming fluidized-bed reactor technology was selected by the U.S. Department of Energy for treatment of sodium-bearing waste at the INL. During the final phase of commissioning prior to radiological operations, regeneration of the high temperature filters in the process gas filter (PGF) vessel became ineffective. After approximately 50 days of testing, the differential pressure across the filters began increasing exponentially and the plant had to be shutdown. Subsequent investigation revealed that the Inconel 625 (I625) sintered metal filter media was corroding in the 550 deg. C reducing environment. A pilot plant run tested four candidate replacement filter media. The top candidate was further tested during a long-term pilot plant run to validate the selection. Data from the testing indicate the selected filters will provide reliable service throughout the life of the project, including after in-situ radiological decontamination. This paper presents the data from this effort and reinforces the value of pilot plant testing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Improvements to the High-Field-Side Transient CHI System on QUEST

Transient coaxial helicity injection (CHI) is a promising candidate for solenoid-free plasma current start-up in a low-aspect-ratio tokamak in support of developing fully non-inductive scenarios. The aim of the transient CHI research on QUEST is to develop a reactor-relevant CHI design. On QUEST, a CHI discharge is initiated by driving current along magnetic field lines that connect an electrically insulated electrode plate (which is referred to as a bias electrode) to a vessel component at the bottom region of the spherical tokamak. In the first application of CHI on QUEST, the electrically insulated electrode plate was biased with respect to the outer vessel in a configuration referred to as low-field-side (LFS) injection. To maintain a narrow injector flux footprint width throughout the discharge, QUEST is now developing a high-field-side (HFS) injection configuration, in which the electrically insulated electrode plate is biased with respect to the inner vessel components. Through the implementation of a CHI-dedicated gas injection system, studies in the HFS injection configuration have now demonstrated good magnetic flux evolution into the vacuum vessel. Toroidal currents up to 43 kA have been achieved, and the generated current has increased in proportion to the magnetic flux connecting the electrodes. Finally, these results which show agreement with the CHI-scaling suggest that much higher levels of toroidal current can be generated on QUEST in an optimized CHI system in which the magnetic flux connecting the CHI electrodes is further increased.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Status Report on ASME Code Development for Nonmetallic Core Components in 2020

The purpose of this report is to provide a status update on the progress and ongoing activities of the current ASME Boiler Pressure Vessel (BPV) Code Section III, Division 5, on nonmetallic core components and assemblies which includes the design rules for graphite and ceramic composite materials. Section III is concerned with the design and construction of nuclear reactor components and Division 5 focuses on high temperature reactors (HTR). The ASME Section III Committee’s function for nuclear items other than pressure-retaining components, is to establish rules of safety related to structural integrity. The term “construction” refers to the all-inclusive effort comprising of materials, design, fabrication, examination, inspection, testing, certification, and pressure relief. The Department of Energy (DOE) supports industry codes and standards development through focused research providing the technical basis for new or modified codes and standards and the participation of subject matter experts in codes and standards committees. The activities for the nonmetallic core components and assemblies committees are the pursuit of two working groups that meet quarterly during ASME BPV Code Week meetings. The two working groups are General Requirements for Graphite and Ceramic Composite Core Components and Assemblies (GR GCCCCA) and Nonmetallic Design and Materials (NDM). The rules on general requirements for nonmetallic core components are discussed in the articles of Subsection HA subpart B. Ceramic composites were recently incorporated within the rules, and the GR GCCCCA working group is now in the process of aligning the requirements with the subsections and articles of ASME BPV Code Section III NCA, the general requirements for Division 1 and 2 of the code. The NDM working group is concerned with the design rules for graphite and composite materials for core components and assemblies discussed in Subsection HH (Class SN Nonmetallic Core Components) subpart A and subpart B which reference graphite and composite materials respectively. The bulk of the progress and work is done within this working group. A significant undertaking, with discussion that started in 2015, is to gain endorsement of ASME Section III, Division 5, by the U.S. Nuclear Regulatory Commission (NRC). In June 2018, ASME formally recommended endorsement to the NRC. The NRC has since initiated the endorsement review process, which is to be concluded early in 2021. ASME task groups on metallic structures and components, as well as non-metallic core support structures, have been formed to define potential pathways and schedules for NRC endorsement of Division 5. NRC determined to perform a review on the 2017 edition of Section III Division 5 as a baseline. As a result, the use of graphite as nonmetallic core components formed part of the review, but it excluded composite materials (which were first introduced in the 2019 edition). This report documents progress in FY 2020 directed toward ASME’s efforts to develop design codes for graphite and composites. Progress includes a major code review by NRC, the first code enquiry, and its solution, an overhaul of the treatment of Weibull Statistics, and the addition of Nonmandatory Appendices to the composites code relating to carbon-carbon composite materials.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Report on the Integration of Experimental and Modeling Data for Initial Equivalence Study of Mechanical Performance in Irradiated LPBF 316 Stainless Steel

To ensure the rapid development, deployment, and use of advanced nuclear technologies, faster qualification approaches are needed. Typically, the primary pathway uses traditional data packages consistent with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, which does not consider the environmental effects the material will experience such as corrosion and radiation damage. Examining radiation effects requires a significant amount of space in US facilities at the Advanced Test Reactor and the High Flux Isotope Reactor (HFIR) and suffers from natural gradients in temperature and neutron flux profiles. Ion irradiation may enable rapid assessment of radiation-induced damage to a material and is proposed as part of an accelerated materials qualification framework through the Advanced Materials and Manufacturing Technologies program. To enhance the utility of ion irradiation as an examination tool, this report provides the initial assessment of engineering-relevant properties of microstructures produced from ion irradiation in the near-surface volume. Nanoindentation, Vickers hardness, and known tensile properties were brought together with simple mathematical models and experimental data for an initial equivalence study of the mechanical performance of irradiated laser powder bed fusion (LPBF) 316 stainless steels across length scales. Direct observation of the calculated ion irradiation yield stress and measured neutron irradiation yield stress at 2 dpa showed that both datasets exhibit the same trend with irradiation temperature and overlap within an acceptable band of stress values. Ion irradiations at 10 dpa serve as a prediction of properties to compare to postirradiation examination of HFIR-irradiated LPBF 316H further in the program. This work is a significant demonstration of the Licensing Approach with Ions and Neutrons, which uses ion irradiations to generate mechanical property information more rapidly than through neutron irradiations.

36 MATERIALS SCIENCE↗