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At least 253 records · Page 14

ASME Section III, Division 5, Class A 100,000-hour design data for Alloy 709

This report documents the 100,000 hour, Class A ASME design material data for Alloy 709 based on the extensive Department of Energy, Office of Nuclear Energy, Advanced Reactor Technologies qualification test campaign. This report includes design tensile properties, creep rupture data, allowable stresses, isochronous stress-strain curves, buckling charts, and a few additional miscellaneous pieces of design data. Companion work at Oak Ridge National Laboratory and Argonne National Laboratory provide design cyclic data --- fatigue charts and creep-fatigue diagrams --- and an inelastic constitutive model. This work substantially completes the ASME data package for the Alloy 709 Code Case, though the design data will continue to updated as the final tests finish. This report also compares the design performance of Alloy 709 against that of 316H stainless steel to provide a reference for the improved high temperature strength of 709 compared to a reference material for sodium fast reactor construction.

36 MATERIALS SCIENCE↗

Numerical study on aerosol sampling in a nuclear facility duct with a 90-degree elbow

Due to the challenging design requirements, elbows are often unavoidable in duct configuration, and these 90-degree bends introduce swirl, velocity variations, recirculation, and secondary flow. These disturbances make it difficult for nuclear facilities to meet particle sampling standards. A series of numerical analyses are conducted to track aerosols in a nuclear facility duct having a 90-degree elbow with the assistance of computational fluid dynamics (CFD). A turbulence model, a continuity, and a momentum, a discrete phase model, and species transport equations are solved simultaneously to track aerosols in the duct. The effect of turbulence models, turbulent dispersion models, droplet drag model, aerosol amount, aerosol spray configuration, guide vanes, and mixers are investigated. Simulation results are analyzed per relevant testing codes such as DOE-HDBK-1169, ASME AG1, ISO 14644-3, ACGIH, and ANSI/HPS N13.1.

Han, Kai [Savannah River Nuclear Solutions (SRNS),↗

Probabilistic Risk Assessment of Aging Layered Pressure Vessels

The National Aeronautics and Space Administration (NASA) operates approximately 300 aging layered pressure vessels that were designed and manufactured prior to ASME Boiler and Pressure Vessel (B&PV) code requirements. In order to make decisions regarding the continued fitness-for-service of these non-code carbon steel vessels, it is necessary to perform a relative risk of failure assessment for each vessel. However, risk assessment of these vessels is confounded by uncertainties and variabilities related to the use of proprietary materials in fabrication, missing construction records, geometric discontinuities, weld residual stresses, and complex service stress gradients in and around the welds. Therefore, a probabilistic framework that can capture these uncertainties and variabilities has been developed to assess the fracture risk of flaws in regions of interest, such as longitudinal and circumferential welds, using the NESSUS® probabilistic modeling software and NASGRO® fracture mechanics software. In this study, the probabilistic framework was used to predict variability in the stress intensity factor associated with different reference flaws located in the head-to-shell circumferential welds of a 4-layer and 14-layer pressure vessel. The probabilistic studies predict variability in flaw behavior and the important uncertain parameters for each reference flaw location.

Riha, D. S.↗

ECAR-6580 Rev 0 ASME Section III, Division 5 Analysis of the MARVEL PCS and GVS Top Corner

The purpose of this Engineering Calculations and Analysis Report (ECAR) is to document the structural evaluation for part of the Microreactor Applications Research Validation and Evaluation (MARVEL) Primary Coolant System (PCS) and Guard Vessel System (GVS). For the PCS this relates specifically to the Distribution Plenum (DP), The Intermediate Heat Exchanger (IHX), and the Upper Downcomer (UD). For the GVS, this will include only the top corner that is machined into the DP Top Plate. These components will be evaluated using the 2021 version of ASME Section III, Division 5 [1] which is a design code that governs the construction of vessels, storage tanks, piping, pumps, valves, supports, core support structures and nonmetallic core components for use in high temperature reactor systems and their supporting systems. Materials at high temperature are subject to creep and fatigue mechanisms that require additional analyses that aren’t covered in ASME Section III, Division 1 rules. Division 5 contains two approaches: elastic or inelastic, however, additional Code Cases specific to Division 5, allow for an Elastic-Perfectly Plastic (EPP) approach. The components analyzed in this ECAR will use a combination of the elastic and EPP approach. Only Design and Service Levels A and B are evaluated in this ECAR. Service Level D evaluations for the entire PCS are documented in ECAR-6564, “MARVEL Project Primary Coolant System Pressure Vessel Stress Documentation” [2] and for the entire GVS are documented in ECAR-6574, “MARVEL Guard Vessel System FEA and ASME Analysis” [3]. The Design and Service Level A and B analyses for the PCS Downcomer piping and Core Barrel are documented in ANL-24/36, "Engineering Calculations and Analysis of the Core Barrel and Downcomer Piping in the MARVEL PCS” [4], and the GVS (except for the top corner) is in ECAR-6574, “MARVEL Guard Vessel System FEA and ASME Analysis.” See Section 2.0 for more detail on the analysis boundaries.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

Vessel Fabrication Overview

LANL utilizes spherical vessels for numerous basic science, materials, and system performance experiments. These pressure vessels are designed and fabricated to meet the intent of ASME B&PV Section VIII, Division 3 and Code Case 2564. LANL has successfully contracted vessel construction in the past, utilizing HSLA-100 steel. As current vessel inventories approach design life, we are once again embarking upon a series of fabrication contracts to produce replacement components and vessel assemblies to support continued experiment cadence at LANL. LANL’s intent is to ramp up a continuous vessel supply chain for the DOE Complex.

42 ENGINEERING↗

A Survey of Variable Extragalactic Sources with XTE's All Sky Monitor (ASM)

The original goal of the project was the near real-time detection of AGN utilizing the SSC 3 of the ASM on XTE which does a deep integration on one 100 square degree region of the sky. While the SSC never performed sufficiently well to allow the success of this goal, the work on the project has led to the development of a new analysis method for coded aperture systems which has now been applied to ASM data for mapping regions near clusters of galaxies such as the Perseus Cluster and the Coma Cluster. Publications are in preparation that describe both the new method and the results from mapping clusters of galaxies.

Jernigan, Garrett↗

Development and performance of INCONEL® alloy 740H® seam-welded piping

INCONEL® alloy 740H® is an age-hardenable nickel-based alloy approved for pressure vessels and piping within ASME Section I, VIII, and B31.1. Currently, the code applies a weld strength reduction factor (WSRF) of 0.7 to the allowable stresses for longitudinal seam welds in the time dependent creep regime. In this work, a full-scale seam weld was successfully produced using typical industrial practices. The component was solution heat-treated and aged after fabrication to improve the performance of the weldment. Tensile, bend, impact, and low-cycle fatigue tests showed the component met all the specification requirements and elevated temperature properties were within base metal expectations with failures predominately in the weld metal. Long-term creep tests, including large samples more representative of the entire weldment, were fabricated and tested to times in excess of 10,000 h at multiple temperatures. Analysis of the creep data supports a WSRF of 0.9 for the solution annealed + aged weldments in contrast to the WSRF of 0.7 currently applied to welded + aged weldments. Metallurgical analysis shows that the solution annealing causes recrystallization of the weld metal and reduces the chemical compositional and microstructural gradients in the weldments resulting in acceptable short-term performance. In long-term creep, evidence for coarsened zones in the weldments due to discontinuous coarsening reactions were identified as the mechanism leading to accelerated creep damage formation in the weld metal and sample failure.

14 SOLAR ENERGY↗

DOE-ART HT(G)R R&D Overview

The US Government currently has no plans to finance or build advanced nuclear reactors. Instead, the DOE performs research and development to raise the technological readiness of new concepts, fuels and component materials so that private industrial partners are more likely to design and build plants. Develop technologies that can enable new concepts and designs to achieve greater levels of safety and resilience, flexibility of use, sustainability and construction or operational affordability. Collaborate with industry to identify and conduct essential research to reduce technical risk associated with advanced reactor technologies. Sustain essential technical expertise and capabilities within national laboratories, universities, and industry to perform needed research. Collaborate with the Nuclear Regulatory Commission and Standards Developing Organizations (ASME, ANS, ANSI, etc.) to address gaps in codes and standards to support advanced reactor designs

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Risk-informed Graded Approach for Reliability and Performance Assessment of Machine Learning and Artificial Intelligence for Advanced Condition Monitoring Techniques

With the shift away from time-based maintenance and toward condition-based maintenance, and to reduce overall maintenance costs, there has been an upsurge in the usage and development of advanced condition monitoring (ACM) techniques for real-time monitoring of nuclear power plant (NPP) components. ACM is particularly useful in the development of digital twins, which are designed to predict the failure or degradation of plant components. Successful implementation of ACM requires an assessment to inform the development of a risk-informed approach to evaluate the use of ACM to meet Nuclear Regulatory Committee (NRC) regulations for in-service testing (IST) programs. This includes the monitoring and diagnostics of reactor components and systems in current, new, and advanced reactors. A key component in ACM is the usage of machine learning (ML) and artificial intelligence (AI) algorithms that can employ real-time data from instrumentation and sensors to detect and predict reactor component degradations. Such predictive capabilities enable early detection of component degradation so as to help plant personnel plan and execute necessary maintenance. For successful implementation of ML/AI in ACM such that regulatory requirements are met, a risk-informed graded approach is needed to assess the reliability and performance of ML/AI for ACM. The American Society for Mechanical Engineers (ASME) developed their Operations and Maintenance (O&M) Code to provide guidance on safe, reliable O&M of NPPs. The IST section of the O&M Code specifically establishes requirements for IST and examination to gauge operational readiness of components in water-cooled NPPs. This paper presents a state-of-the-art review of how reliability and risk assessment can be integrated with ACM to assess component performance by non-nuclear industries. This is followed by different methodologies and approaches for conducting performance and reliability assessments so as to meet IST requirements for NPP components.

97 - MATHEMATICS AND COMPUTING↗

Risk-informed Graded Approach for Reliability and Performance Assessment for Advanced Condition Monitoring Techniques

With the shift away from time-based maintenance and toward condition-based maintenance, and to reduce overall maintenance costs, there has been an upsurge in the usage and development of advanced condition monitoring (ACM) techniques for real-time monitoring of nuclear power plant (NPP) components. ACM is particularly useful in the development of digital twins, which are designed to predict the failure or degradation of plant components. Successful implementation of ACM requires an assessment to inform the development of a risk-informed approach to evaluate the use of ACM to meet Nuclear Regulatory Committee (NRC) regulations for in-service testing (IST) programs. This includes the monitoring and diagnostics of reactor components and systems in current, new, and advanced reactors. A key component in ACM is the usage of machine learning (ML) and artificial intelligence (AI) algorithms that can employ real-time data from instrumentation and sensors to detect and predict reactor component degradations. Such predictive capabilities enable early detection of component degradation so as to help plant personnel plan and execute necessary maintenance. For successful implementation of ML/AI in ACM such that regulatory requirements are met, a risk-informed graded approach is needed to assess the reliability and performance of ML/AI for ACM. The American Society for Mechanical Engineers (ASME) developed their Operations and Maintenance (O&M) Code to provide guidance on safe, reliable O&M of NPPs. The IST section of the O&M Code specifically establishes requirements for IST and examination to gauge operational readiness of components in water-cooled NPPs. This paper presents a state-of-the-art review of how reliability and risk assessment can be integrated with ACM to assess component performance by non-nuclear industries. This is followed by different methodologies and approaches for conducting performance and reliability assessments so as to meet IST requirements for NPP components.

99 - GENERAL AND MISCELLANEOUS↗

HSLA-100 Forging Contract Vendor Post-Award Meeting [Slides]

LANL utilizes thick-walled, spherical vessels for numerous basic science, materials, and system performance experiments. These pressure retaining vessels are designed and fabricated to meet the intent of ASM E B&PV Section VIII, Division 3 and Code Case 2564. LANL has successfully contracted past production of component parts (nozzle and cover forgings, plate production and head forming) and overall vessel construction (welding, inspection, hydrostatic testing) utilizing HSLA-100 steel. As current vessel inventories approach design lifetimes, we are once again embarking upon a series of fabrication contracts to produce replacement components and vessel assemblies to ensure continued experiment cadence at LANL. We anticipate releasing additional future vessel production contracts as experimental demands at LANL continue to increase

36 MATERIALS SCIENCE↗

Design Basis Document / Owner’s Technical Specification for Nitrate Salt Systems in CSP Projects (Final Technical Report)

The number of commercial coal, gas, and nuclear projects built over the past 100 years number in the thousands. As such, there is a large database, available to a wide range of commercial engineering contractors, on proven designs. In essence, unsuccessful designs have, through generations of iterations, been identified and then deleted from further consideration. In contrast, the number of commercial parabolic trough projects using nitrate salt for the thermal storage media is perhaps 60. Further, the number of commercial central receiver projects using nitrate salt as the working fluid is on the order of 20, including estimates for China. Given the relative immaturity of salt technology, and commercial pressures to successfully bid new solar projects into a mature electricity market, solar projects often promise more than has been delivered. The Design Basis Document / Owner’s Technical Specification is a first step in the iteration process. The report describes the successful features of commercial projects, outlines a range equipment and system failures in projects that didn’t operate as intended, and provides a draft set of design changes intended to correct the known problems. The product of the study is 3 volumes of technical material; one volume is on parabolic trough technologies; a second is on central receiver technologies, and the third is on potential design changes to parabolic trough and central receiver projects. The 3 volumes, which total some 590 pages, can be found at https://www.solardynllc.com/csp-plant-technologies. One of the principal topics in the report is the use of functional or prescriptive specifications. Functional specifications describe what the equipment needs to do, consistent with the minimum legal requirements of the local jurisdictions. The details of how this is to be accomplished is developed by the engineering contractor. Prescriptive specifications, which are developed by the Owner, prescribe to the engineering contractor how the functional requirements are to be met. This arrangement ensures that the favorable experience from a previous project is repeated. One example is the design code for the hot salt tank in central receiver projects. The closest design basis is API Standard 650 Welded Steel Tanks for Oil Storage. However, the maximum design temperature in API 650 is 260 °C. As such, solar projects have typically adopted a hybrid Code approach, in which allowable material stresses are taken from ASME Section II Materials. Further, since the tanks experience daily changes in temperature and in (static) pressure, and since portions of the tank can operate at stresses beyond the elastic range, the low cycle fatigue life of the tank is conducted using the rules of Section VIII Division 2. However, in a recent study by NREL, the principal damage mechanism was identified as creep rather than fatigue. Further, design stresses permitted under Section VIII Division 2, corresponding to a fatigue life of 30 years, result in projected creep lifetimes of only 2 to 5 years. An alternate design approach, prescribed by the Owner, would be based on Code sections intended for high temperature service in the creep regime. A candidate is Section III Division 5. Granted, this is a nuclear code section, and it’s use would not likely be mandated by local jurisdictions. However, the effects of creep have been deemed to be of sufficient importance that one nuclear project developer, and one central receiver project developer, have stipulated in the tank design specification that the equipment be designed to the requirements of Section III Division 5.

14 SOLAR ENERGY↗

SWIFT Code Assessment for Two Similar Transonic Compressors

One goal of the NASA Fundamental Aeronautics Program is the assessment of computational fluid dynamic (CFD) codes used for the design and analysis of many aerospace systems. This paper describes the assessment of the SWIFT turbomachinery analysis code for two similar transonic compressors, NASA rotor 37 and stage 35. The two rotors have identical blade profiles on the front, transonic half of the blade but rotor 37 has more camber aft of the shock. Thus the two rotors have the same shock structure and choking flow but rotor 37 produces a higher pressure ratio. The two compressors and experimental data are described here briefly. Rotor 37 was also used for test cases organized by ASME, IGTI, and AGARD in 1994-1998. Most of the participating codes over predicted pressure and temperature ratios, and failed to predict certain features of the downstream flowfield. Since then the AUSM+ upwind scheme and the k- turbulence model have been added to SWIFT. In this work the new capabilities were assessed for the two compressors. Comparisons were made with overall performance maps and spanwise profiles of several aerodynamic parameters. The results for rotor 37 were in much better agreement with the experimental data than the original blind test case results although there were still some discrepancies. The results for stage 35 were in very good agreement with the data. The results for rotor 37 were very sensitive to turbulence model parameters but the results for stage 35 were not. Comparison of the rotor solutions showed that the main difference between the two rotors was not blade camber as expected, but shock/boundary layer interaction on the casing.

Chima, Rodrick V.↗

ASME BPVC Section III Division 5 Application

Using AMSE assessments Simplified assessment overview Implementation Limitation and important considerations Example problems Full Assessment overview Implementation Limitation and important considerations Example problems Limitation of the code

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

GCR: Class B Code Case

This PowerPoint presentation is for Joint ART Materials/AMMT Program Review, Germantown, MD, on June-5-8, 2023. It will discuss development to revamp the ASME Section III Division 5 Class B design rules. There will be a plan in the FY24 to assess extrapolation methods for other Class A materials, alternative strain range evaluations for fatigue damage, evaluation of Class B intersection point in D-diagram relative to material-specific Class A intersection points, and evaluation of new Class B rules against Class A rules based on Elastic-Perfectly Plastic (EPP) methodology and full inelastic analysis method, using sample problems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

FY25 Status Update on A709 Code Case Testing at ANL

This report provides an update on the status of the creep rupture testing on the precipitation-treated (PT) Alloy 709 samples fabricated from the first, the second and the third commercial heats, in support of the first, 100,000-hour Alloy 709 Code Case development under the American Society of Mechanical Engineers (ASME) Section III, Division 5 for high-temperature reactor construction. In Fiscal Year (FY) 25, no new tests were initiated, while 5 previously loaded tests were completed. This report presents an overview of the creep-rupture data gathered at Argonne National Laboratory (ANL) since FY21 that totaled 52 completed tests. Metallographic images showing crack morphologies in some representative samples ruptured in FY25 are also included. Key conclusions include: 1) the creep behavior of the three heats of Alloy 709 can be well described by the Larson-Miller relationship; 2) higher temperatures and larger stresses lead to shorter creep life and higher minimum creep rate; 3) Alloy 709 is overall very ductile under creep; 4) the crack morphologies near fracture surfaces are dependent on material’s grain size and test conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Application of GOTHIC to Groundwater Transport Analysis - 20152

Migration of pollutants and hazardous wastes, potentially containing radioactive isotopes, via groundwater transport is a concern at most waste cleanup sites. Predictive analysis can be used to evaluate mitigating actions intended to minimize impact on the environment and public exposure. GOTHIC is a multipurpose thermal hydraulics code that is used extensively in the nuclear industry for design, licensing and operation evaluations. It combines the capabilities of typical one-dimensional system codes and the essential features of Computational Fluid Dynamics (CFD) codes for three-dimensional analysis. There are other codes that are specifically developed for groundwater transport analysis and the results presented here are consistent with prior analyses. However, GOTHIC has some unique features that offer advantages for applications related to nuclear waste. Most importantly, it has been developed and maintained under a Quality Assurance program in compliance with the requirements of 10CFR50 Appendix B [2] and applicable portions of ASME NQA-1 [3] since 1995. Available GOTHIC capabilities that make the code especially useful for groundwater transport of nuclear materials include: - Tracking of any number of tracer elements for contaminants and other species of interest; - Radioactive decay and progeny of tracer elements; - Adsorption/desorption of tracer elements; - Tracking of any number of dissolved gases; - Release and absorption for dissolved gases; - Vapor phase tracking; - Non-Newtonian fluid modeling. The general porous body modeling approach makes GOTHIC well suited to groundwater transport analysis. The multi-region modeling approach used by GOTHIC simplifies model construction for regions of varying hydrologic characteristics and focuses the computational effort on regions of particular interest while simultaneously capturing the macroscopic response and any feedback effects across the larger domain. The applicability of GOTHIC to groundwater transport applications is demonstrated by comparing code results with available analytic or semi-analytic solutions for groundwater behavior. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗