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All sources NASA NTRS DOE OSTI

DOE OSTI2026

Development of a Framework and Methodology for an Advanced Reactor Materials Environmental Effects Design Guide

Advanced non-light-water reactor components may operate at elevated temperature while experiencing cyclic loading, significant neutron irradiation, and exposure to reactor coolant. ASME Boiler and Pressure Vessel Code, Section III, Division 5, provides design rules for elevated-temperature service but does not include specific procedures to account for environmental effects on material properties. This report develops an initial framework and methodology for an Environmental Effects Design Guide (EEDG) focused on neutron irradiation; coolant-environment effects are reserved for future work. The proposed approach treats irradiation as a property-based overlay on the existing Division 5 design process, with two routes: a sparse-data route applying two reduction factors — FCR on creep-rupture strength and FF on fatigue life — for the creep-fatigue evaluations that typically control the design of advanced high-temperature reactor components, and a fuller framework developing the property-to-rule chain across the four Division 5 checks (primary load, strain limits and ratcheting, creep-fatigue, and buckling), together with swelling and weldments as scope items. Both routes are scoped by an in-pile qualification that restricts the use of post-irradiation-examination-derived properties in regimes where an in-pile mechanism could control the design outcome. Illustrative outputs derived on a compiled annealed Type 316 database — FCR ≈ 0.78–0.86 and FF ≈ 0.4 — demonstrate the calculation method within that specific dataset. The framework is an initial, testable design-rule concept; it identifies a practical path for preliminary design evaluations under sparse data and the material data and testing needed to develop the framework further.

Barua, Bipul (ORCID:0000000247184113)↗

DOE OSTI2024

ASME – Grappling with the Concept of Component Failure

ASME concepts of component failure. Definition for failure, functionality, damage tolerance, including Magnox reactors and AGRs. How ASME is addressing damage tolerance, and Section III division 5 Design, Classification, Reliability Target for SRC-1 components, Basic analysis approach, simple assessment, and full assessments, Reference to reliability and integrity management and strategy for graphite components.

36 MATERIALS SCIENCE↗

DOE OSTI2024

ASME Design Code Rule Changes for Nuclear Graphite

The American Society of Mechanical Engineers Boiler Pressure and Vessel Code (ASME BPVC) Section III, Division 5, Article HHA-3000 outlines graphite core component and graphite core assembly design guidelines. Graphite core components are defined as ?components manufactured from graphite that are installed to form a graphite core assembly within the reactor pressure vessel of a high temperature, graphite moderated fission reactor.? (p. 413) Graphites? inherent defect distributions do not allow for deterministic material reliability. Rather, graphite has variable strength distributions which change by grade. Article HHA-3000 outlines two semi-probabilistic methods, the full and simplified assessments, which set design load limit targets for each of three component structural reliability classes. The Design Task Group was officially recognized as a specialized task group within ASME November of 2023, though we?ve been collaborating since 2022. The purpose of the Design Task Group is to correct, clarify, and make HHA-3000 function as intended. The Design Task Group will sunset once we?ve achieved our objectives. The Design Task Group was specifically told to not write new Code. While there may be more precise and more accurate methods to determine reliability targets, the current methods are conservative, relatively simple to implement, and have thus far been considered satisfactory for setting design reliability targets. Much of the ground-work to write proposal files and background documents for records to make the changes needed to achieve our objective have been completed. The Design Task Group has documented much of their work through papers, presentations, and memorandums. Three memorandums in which INL team members had substantial contributions are found in the Appendices: FEA Modeling for the Baseline Program, Evaluating the Effects on Margin of Updating the Threshold and Shape Parameters in the Full Assessment, and Interpretations of the Full and Simplified Assessments in ASME BPVC. Most of the on-going work to achieve the Design Task Group?s objective will be addressing comments on existing records and moving records through the balloting process. The Design Task Group met bi-weekly mostly through the end of FY2023. Since February 2024, the Design Task Group has mostly been completed with solving and documenting the technical issues associated with the assessments. Unless new tasks are identified, the remaining work of the Design Task Group will be political and editorial.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2024

FY24 progress report on A709 mechanical properties data development and A709 thermal aging status

The report provides the status of the creep, fatigue, and creep-fatigue testing to date conducted at Idaho National Laboratory to generate the data package. This data package evaluates the material performance from three commercial heats to support the Alloy 709 qualification in American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 5. First procured heat was manufactured by G.O. Carlson with heat number 58776. Second and third heats were fabricated by ATI Specialty Rolled Products with heat numbers 529900 and 530843, respectively. A series of creep and cyclic specimens were fabricated from these three heats and tests were performed. A list of finished, ongoing, and planned test are presented for creep, fatigue, and creep-fatigue tests. Cyclic properties of three commercial heats are compared.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Creep-Fatigue Properties of Additional 316H PM-HIP Materials Fabricated from Different Powder Compositions and Processing Routes

The process of powder metallurgy (PM) hot isostatic pressing (HIP) works by consolidating powdered materials at relatively high temperature and pressure to form near-net-shaped components. Ideally, PM-HIP production methods can reduce component lead time and improve designs for high-temperature reactors and/or microreactors. To introduce PM-HIP into Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, it is necessary to show adequate material properties regarding creep, high-temperature low-cycle fatigue, and creep fatigue. However, prior work has shown that the creep-fatigue cycles to failure for PM-HIP 316H stainless steel are greatly reduced compared to the conventional, wrought product. This work continued creep-fatigue analysis on a 316H stainless steel with lower oxygen and nitrogen contents and at different HIP parameters than previously analyzed. The objective was to better understand what is causing the reduced PM-HIP 316H performance so improvements can be made PM-HIP 316H creep-fatigue lifetimes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Evaluation of the Simplified Assessment Peak Equivalent Stress Design Limit Probability of Failure

The ASME Boiler and Pressure Vessel Code Section III Division 5 Subsection HH Subpart A (HHA) outlines requirements and qualification methods for graphite core components in high temperature reactor environments. One such structural qualification method is the Simplified Assessment outlined in HHA-3220, which represents a comparison of service level specific equivalent stresses to allowable stress values derived from target probability of failures (POFs). The target POF is based on the structural reliability class (SRC) for the components and service condition of interest, and is used to establish design limits for combined membrane and peak equivalent stresses. Combined membrane stress limits are derived from two parameter Weibull distribution fits of the graphite grade specific tensile strength data and the target POF. The combined membrane stress limits are then scaled based on the ratio of flexural strength to tensile strength (referred hereinafter as Rtf) to establish the corresponding peak equivalent stress limits. It is desired that the target POF is at least maintained for both combined membrane and peak equivalent design limits, especially with the use of Rtf. The assessment presented herein examines tensile and flexural data for medium and fine grade graphites relative to the calculated combined membrane and peak equivalent stress limits per HHA-3220 and illustrates the simplified assessment method results in anticipated, or at least conservative, peak equivalent design stress limits.

36 - MATERIALS SCIENCE↗

DOE OSTI2025

Summary report on the mechanical testing performed at INL for creep rupture, fatigue, creep-fatigue, and aging of A709 base material

The report provides the status of the creep, fatigue, and creep-fatigue testing to conducted at Idaho National Laboratory to generate a data package to support Alloy 709 qualification in American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 5. These tests were performed on three commercial heats of plate material. The first procured heat was manufactured by G.O. Carlson with heat number 58776. Second and third heats were fabricated by ATI Specialty Rolled Products, with heat numbers 529900 and 530843, respectively. A series of creep and cyclic specimens were fabricated from these three heats, and tests were performed. This report presents a master list of all completed and ongoing creep, fatigue, and creep fatigue tests. Cyclic properties of three commercial heats are compared. Thermal aging of three commercial heats for 3000 hours were completed in FY 2025. The thermal aging campaign to age plate material from three commercial heats up to 100,000 hours is ongoing.

36 - MATERIALS SCIENCE↗

DOE OSTI2025

FY24 progress report on A709 mechanical properties data development and A709 thermal aging status

36 - MATERIALS SCIENCE↗

DOE OSTI2024

ECAR-6574 Rev 1 MARVEL Guard Vessel System FEA and ASME Analysis

This ECAR encompasses the design-by-equation requirements in ASME BPVC Section III Division 5 Subsection HCB, in addition to the service level and design-by-analysis requirements of Subsection HBB, as they apply to the Guard Vessel.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2024

High Temperature Materials

Overview of high temperature metals, including component construction rules for advanced reactor designs, assessment of filler metal for Alloy 800H weldments, high-temperature crack growth rate testing, work packages and contributors for FY24, creep-fatigue damage summation evaluation, GCR metal needs, and high temperature metals from other programs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2014

Erratum To: Indigenous Frameworks for Observing and Responding to Climate Change in Alaska

In section 5, item 1 of this article it is stated that: A recent shift in decision-making authority from the politically appointed Board of Game to the Subsistence Division of the Alaska Department of Fish and Game should make these decisions about hunting regulations more responsive to local observations and needs. We now recognize that this shift in regulatory authority to ADF&G never occurred. We hereby correct this error so that wildlife users in Alaska do not come to ADF&G with expectations that this agency has authority to adjust hunting regulations to accommodate climate change.

Alaska↗