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DOE OSTI2016

Target Housing Material Options

With gas cooling, heat transfer coefficients are low compared to water. The benefit of gas from a heat transfer point of view is that there is really no upper temperature limit for the coolant, as compared to water, which is limited ultimately by the critical point, and in practice the critical heat flux. In our case with parallel flow channels, water is limited to even lower operating limits by nucleate boiling. So gas can get as hot as the containment material will allow, but to get the density and heat transfer up to something reasonable, we must also increase pressure, thus increasing stress on the containment, namely the front and back faces. We are designing to ASME BPVC, which, for most materials allows a maximum stress of UTS/3. So we want the highest possible UTS. For reference, the front face stress in the 12 mm target at 300 psi was about 90 MPa. The inconel 718 allowable stress at 900°C is 1/3 of 517 or 172 MPa. So we are in a very safe place, but the uTS is dropping rapidly with temperature above 900°C. As we increase target diameter, the challenge will be to keep the stress down. We are probably looking at keeping the allowable at or above the present value, and at as high a temperature as possible.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2018

Fundamental Understanding of Creep-Fatigue Interactions in 9Cr-1MoV Steel Welds. Final Report

Ferritic-martensitic steels constitute a vital class of materials for current and future generation of nuclear reactors as well as ultra supercritical coal fired power plants due to their superior creep resistant properties. One particular ferritic-martensitic steel, 9Cr-1MoV (Grade 91) is one of five materials approved for high temperature use in nuclear service by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC). Grade 91 is also a candidate material for proposed fission and fusion reactors such as the sodium cooled fast reactor (SFR). In service, Grade 91 components can be subjected to elevated temperatures and cyclic stresses, leading to the accumulation of creep-fatigue (CF) damage. While the monotonic creep properties of Grade 91 steels have been studied extensively, there is less knowledge concerning the mechanisms governing CF damage of this steel in the creep-dominated failure region. Moreover, the creep failure is especially pronounced in welded Grade 91 components, which are commonly joined using conventional arc welding techniques such as flux-cored arc welding (FCAW). In a Grade 91 weld, the creep failure typically occurs in the outer region of the heat affected zone (HAZ), specifically in the intercritical or fine-grained heat-affected zone (ICHAZ or FGHAZ), a phenomenon known as type IV failure. Currently, there exists a limited knowledge of CF damage mechanisms in Grade 91 welds. Improving the mechanistic understanding of the CF damage and failure of Grade 91 steel and its welds with inhomogeneous microstructure is crucial for the safety and design of key structural components in advanced, high-temperature reactors. The overall objectives of the project were to advance the state of knowledge and understanding of CF damage in both Grade 91 base metal and its welds, especially under loading conditions where creep is the dominant damage mechanism. Uncovering the mechanisms governing the creep-fatigue deformation and damage in Grade 91 base metal in creep-dominated region as well as type IV failure in Grade 91 welds contributes to a new knowledge that may lead to more accurate life prediction models than those currently in use. Moreover, the superior CF properties obtained by low heat input welding has a high potential to develop practical solutions to the premature failure of Grade 91 weld.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Report on the ASME design code development effort for Composite Core Components for High Temperature Reactors

This report summarizes the activities and status of the ASME design code development effort for Composite Core Components for High Temperature Reactors during Fiscal Year 2018. The subsection drafts develop for ASME BPVC Division 5 Section III, namely Subsection HA Subpart B and Subsection HH Subpart B, are near publication. It is likely that Subsection HH Subpart B will be accepted during the next ASME review meeting. Subsection HA Subpart B may have to continue through more review iterations before final acceptance. Two new standards for fiber-reinforced advanced ceramic composites have been published and a report on the technology gap for Molten Salt Reactor applications of SiC/SiC composites was completed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Historical Context and Perspective on Allowable Stresses and Design Parameters in ASME Section III, Division 5, Subsection HB, Subpart B

In the 1970s, the U.S. Nuclear Regulatory Commission (NRC) issued Regulatory Guide (RG) 1.87 (Rev 1, June 1975) to provide interim licensing guidelines to aid applicants in implementing the 10 CFR Part 50 requirements with respect to American Society of Mechanical Engineers (ASME) Boiler Pressure and Vessel Code (BPVC) Class 1 components operating at elevated temperatures for high-temperature gas-cooled reactors, liquid-metal fast-breeder reactors, and gas-cooled fast-breeder reactors. RG 1.87 referenced ASME Code Case 1592 for materials and design and Code Cases 1593, 1594, 1595, and 1596 for fabrication and installation, examination, testing, and overpressure protection, respectively.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Equivalent Safety Evaluation of the 2021 Edition of the ASME Boiler and Pressure Vessel Code to Editions Incorporated by Reference in 49 CFR 192, 193, and 195

Federal safety standards for natural gas and hazardous liquid pipelines and liquid natural gas facilities incorporate rules and requirements for boiler and pressure vessel design and fabrication into 49 CFR Parts 192, 193, and 195 through the IBR process. The equivalent safety evaluations documented in this report demonstrate that boilers and pressure vessels in pipeline facilities that are designed and fabricated in accordance with rules and requirements specified in the 2021 edition do not violate the fundamental safety assumptions stated or implied in the 2007 edition of the ASME BPVC and therefore provide equivalent safety.

03 NATURAL GAS↗

DOE OSTI2023

Design and Scoping Tests on Alloy 617 Using Notched Specimen Geometry to Validate Methods for Multiaxial Stress Relaxation

In FY 2022, a development effort was initiated at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) to examine the multiaxial stress-relaxation behavior and multiaxial stress-state effect on the creep-fatigue (CF) performance for Alloy 617 at elevated temperatures. This effort supported the development of the design rules in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 5. In this work, two types of the notch specimen geometries were designed. An inelastic constitutive model was used to investigate the stress triaxiality and elastic follow-up effects on the notch specimens under CF deformation in the finite element (FE) simulations. The numerical study demonstrated that the stress triaxiality and elastic follow-up caused by the notches both play significant roles in the stress-relaxation behavior. In addition, Alloy 617 CF experiments were designed, and the testing on the notch specimens is ongoing. CF test failure data will be generated on the specimens with notches under various conditions in FY 2023, and the results will be used to validate methods for accounting for the multiaxial stress relaxation effect in the design code.

36 MATERIALS SCIENCE↗

DOE OSTI2023

An Assessment of Machine Learning Applied to Ultrasonic Nondestructive Evaluation

In the United States, the nuclear industry performs inservice inspection (ISI) through nondestructive examination (NDE) methods in accordance with guidelines specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Ultrasonic nondestructive testing and evaluation (NDT&E) is one of the more commonly used techniques for inspecting Class 1 structural components in nuclear power systems. As the number of qualified NDE inspectors declines, the nuclear industry is looking to take advantage of advances in automation to enhance inspection capabilities. Advances in computational power, cloud-based computing, and machine learning algorithms make automated data analysis possible. Machine learning (ML) has shown huge potential in automated data analyses for ultrasonic NDE in the context of weld inspections.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Fitness For Service Assessment of a Corroded Heat Exchanger

Within the Fermi National Accelerator complex, there exist various water systems that support accelerator operations. One of these systems is extremely vital to the operation of the machine; that is the cooling system. The cooling system consists of nine relatively large heat exchangers that take untreated pond water and use it to cool the process fluid that further cools machine components. Over the 30 years these heat exchangers have been in operation, they have undergone significant material loss on the channels. This material loss, due to various forms of corrosion such as galvanic and microbiologically influenced corrosion (MIC) and possibly others, has deteriorated more than 80% of the nominal wall thickness of some of the exchangers and placed them in a questionable state. ASME FFS-1 (API 579) has been applied to address the condition of the heat exchangers due to their noncompliance with the governing code, BPVC Sec. VIII Div. 1. The assessments encompassed ASME FFS-1 parts 4: General Metal Loss and 9: Crack Like Flaw using level 1, 2, and 3 analysis techniques based on inspection data obtained by API 510 inspections. Level 1 and 2 assessments were deemed unfit for the corroded regions due to their location relative to a major structural discontinuity (channel to tube-sheet joint), so a level 3 analysis was conducted according to ASME Sec. VIII Div. 2 (design by analysis) rules for pressure vessels. Supplemental information included pond water tests to determine an accurate future corrosion allowance due to lacking inspection history. A leak before break (LBB) route was chosen to evaluate the possibility of leaking prior to the onset of failure. The analysis of one heat exchanger shows that the possibility the channel will develop a pinhole leak over 2.5 more years of operation should not be overlooked, but burst was unlikely from operation. The use of fracture mechanics show, that if a through-wall crack were to develop, it would not propagate further than the channel geometry and cause a leak not greater than 35 GPM. Using ASME Section XI Code Case N-705-1, allowing us to operate with a leak until the next outage given certain operating conditions and developing a leak mitigation procedure, this heat exchanger is deemed fit-for-service.

Humenik, Alex [Fermilab]↗

DOE OSTI2025

Microstructure-Based Understanding of High-Temperature Deformation Behaviors in Laser Powder Bed Fusion (LPBF) 316H Stainless Steel

This report presents the results of high-temperature mechanical property testing and microstructural analysis of laser powder bed fusion (LPBF) 316H stainless steel (SS) conducted during Fiscal Year 2025 (FY25) under the U.S. Department of Energy, Office of Nuclear Energy’s Advanced Materials and Manufacturing Technologies (AMMT) program. The study builds upon prior work by expanding the mechanical property test matrix and advancing a microstructure-based mechanistic understanding of LPBF 316H SS performance, with a focus on the solution annealed (SA, 1100°C for 1 hour) materials. High-temperature tension, creep, fatigue, and creep-fatigue tests were performed. Thermal aging studies were conducted. Advanced characterization techniques, including scanning electron microscopy (SEM), scanning transmission electron microscopy (STEM), and high-energy synchrotron X-ray diffraction, were employed to investigate the microstructural evolution during thermal aging and mechanical testing. It was discovered that the as-built (AB) LPBF 316H SS exhibited precipitation kinetics during thermal aging that are 10–100 times faster than its wrought counterpart due to the presence of high densities of preferred nucleation sites, while the precipitation kinetics in SA LPBF 316H was under investigation. The accelerated precipitation of embrittling phases such as sigma phase in LPBF 316H SS compared to its wrought counterpart led to significant impacts on creep ductility. The influence of laser printing parameters and build orientations on fatigue and creep-fatigue performance was observed. In assessing the creep-fatigue performance, it was discovered that the SA LPBF 316H samples failed in less than 200 cycles under 595°C, 0.5% (strain amplitude) with a 60-minute hold time. This work establishes a mechanistic framework for predicting the long-term behavior of LPBF 316H SS, supporting its rapid qualification under the ASME BPVC. The findings contribute to the broader goal of enabling the deployment of advanced materials in next-generation nuclear energy systems. An outlook for FY26 work, including continued aging studies, mechanical testing, and microstructural characterization, is provided.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Initial Design Curves for Alloy 709 for an Improved Creep-fatigue Design Method

Creep-fatigue (CF) interaction damage is the primary damage mode for high-temperature structural components subjected to cyclic loading. Over the past several decades, researchers within the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 5, have focused on developing elevated temperature code rules to ensure conservative structural designs that mitigate CF failure in high-temperature reactors. The existing CF evaluation methodologies in the Code are based on the creep and fatigue damage diagram approach, which is complex and often excessively conservative. The alternative CF evaluation approach proposed here is intended to significantly simplify the evaluation procedure while reducing conservatism in high-temperature component design analysis. This alternative CF evaluation method integrates the elastic–perfectly plastic (EPP) analysis approach with the simplified model test (SMT) CF design concept, leveraging the advantages of both methods. This report presents the preliminary analysis and the approach for developing CF design curves for Alloy 709, utilizing fatigue and CF data generated for the 100,000-hr Code Case to support its qualification to ASME Section III, Division 5 for Class A construction of high temperature reactors. This study is to support the incorporation of Alloy 709 in this alternative CF evaluation method. Recommendations for the remaining work needed to complete the effort are also provided.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Evaluation of the Mechanical Properties of Cast and Wrought CF8C-Plus Relevant to ASME Code Case Qualification

The mechanical behavior of a cast form of an advanced austenitic stainless steel, CF8C-Plus, is compared with that of its wrought equivalent in terms of both tensile and creep-rupture properties and estimated allowable stress values for pressurized service at temperatures up to about 850°C. A traditional Larson-Miller parametric model is used to analyze the creep-rupture data and to predict long-term lifetimes for comparison of the two alloy types. The cast CF8C-Plus exhibited lower yield and tensile strengths, but higher creep strength compared to its wrought counterpart. Two welding methods, shielded-metal-arc welding (SMAW) and gas-metal-arc welding, met the weld qualification acceptance criteria in ASME BPVC Section IX for the cast CF8C-Plus. However, for the wrought CF8C-Plus, while SMAW and gas-tungsten-arc welding passed the tensile acceptance criteria, they failed the side bend tests due to lack of fusion or weld metal discontinuities.

Chen, Xiang↗

DOE OSTI2020

Alloy 709 Materials Procurement

Qualifying a new material for design and construction of elevated temperature nuclear components under Section III Division 5 of the BPVC requires measurement of a variety of properties, including tensile, creep, fatigue and creep-fatigue behavior in the relevant temperature range, and from the relevant product form, from a commercial scale heat. For the ART program the desired product form is hot rolled plate. This procurement is for a unique composition of stainless steel for which a commercial scale heat represents melting about 50,000 pounds of the alloy, resulting in about 45,000 pounds of usable plate. The Alloy 709 plate from this purchase will be provided to INL, Argonne National Laboratory and Oak Ridge National Laboratory for characterization of the relevant properties to support Code qualification.