Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “ASME”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 559 records · Page 31

V31 Test Report

The V31 containment vessel was procured by the US Army Recovered Chemical Material Directorate (RCMD) as a third - generation EDS containment vessel. It is the fifth EDS vessel to be fabricated under Code Case 2564 of the 2019 ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel, based on the code case, is twenty-four (24) pounds TNT - equivalent for up to 1092 detonations. This report documents the results of explosive tests that were performed on the vessel at Sandia National Laboratories in Albuquerque, New Mexico to qualify the vessel for field operations use. There were three design basis configurations for qualification testing. Qualification test (1) consisted of a simulated M55 rocket motor and warhead assembly of 24lbs of Composition C-4 (30 lb TNT equivalent). This test was considered the maximum load case, based on modeling and simulation methods performed by Sandia prior to the vessel design phase. Qualification test (2) consisted of a regular, right circular cylinder, unitary charge, located central to the vessel interior of 19.2 lb of Composition C-4 (24 lb TNT equivalent). Qualification test (3) consisted of a 12-pack of regular, right circular cylinders of 2 lb each, distributed evenly inside the vessel (totaling 19.2 lb of C-4, or 24 lb TNT equivalent). All vessel acceptance criteria were met.

42 ENGINEERING↗

Innovative Method for Welding in Generation 3 CSP to Enable Reliable Manufacturing of Solar Receivers to withstand Daily Cycling at Temperatures Above 700°C (Final Technical Report)

Inconel® Alloy 740H® (alloy 740H) was the first age-hardenable nickel-based alloy approved by the ASME Boiler & Pressure Vessel Code for use in pressure-boundary applications. Over the past ~20 years the alloy has been optimized for weldability and high-temperature stability, approved for use in different applications. Development of a supply chain combined with the advantageous properties of the alloy (high-temperature creep strength, oxidation and corrosion resistance, etc.) have resulted in the alloy being applied to new high-temperature power cycle demonstration projects, and of particular interest are applications to concentrating solar power (CSP) to enable higher-efficiency Generation 3 CSP systems and the corresponding supercritical CO 2 (sCO 2 ) power cycle components (heat exchangers, piping, etc.). The high allowable stresses of alloy 740H also make it a desirable material for current Generation 2 CSP solar power receivers to improve cyclic capability and/or reduce receiver height. Recent experiences in demonstration projects utilizing alloy 740H identified cracking issues during welding and fabrication. In this project, a detailed study was done to confirm and clarify the Stress Relaxation Cracking (SRxC) mechanism, also known as stress relief cracking or strain-age cracking (SAC), during post-weld heat-treatment (PWHT). This involved detailed microscopy and advanced characterization to understand the root cause(s) of three failures obtained from industry. Based in-part on these findings, a targeted laboratory based SRxC test method was utilized to evaluate variables such as heat-to-heat variations, strain level, PWHT temperature, and starting material condition on three heats of alloy 740H. Industrial shop welding of cold-worked plates was also conducted. The research showed the following: SRxC was confirmed as the cracking mechanism for all field failures; Stress state (from residual stresses, constraint, deformation, and local stress concentrations) was playing a significant role in field failures and laboratory testing confirmed increasing susceptibility for all heats with increasing strain levels. High levels of microstructural strain were identified at crack initiation locations, in some cases leading to local recrystallization; Precipitate free zones (PFZs) at grain boundaries were found at relaxation cracks and crack initiation locations uniquely associated with SRxC in alloy 740H. Laboratory testing reproduced this microstructural feature which had only previously been reported in long-term creep testing of weldments. Advanced nano-scale characterization confirmed the presence of a moving boundary leading to coarsening of precipitates and PFZs where damage accumulated; The research suggested heat-to-heat variations due to local chemistry and processing may influence SRxC susceptibility, but more work is needed to fully clarify these effects. To disseminate the key learnings from this research to the scientific and engineering communities and alloy 740H end users, multiple technical publications and presentations were made, an industrial alloy 740H users meeting was held, and a new industry guideline specification document which can be directly implemented by end-users of alloy 740H was produced.

14 SOLAR ENERGY↗

H-Canyon Jumper Gasket Sealing Evaluation

H-Canyon Documented Safety Analysis approved December 2019 credits the leak integrity of H-Canyon (HCA) process tanks, piping, and jumpers in preventing the release of radioactive material in a post seismic environment. Demonstration of leak tightness capability for this equipment was achieved through substantial analytical effort. However, one potential weak point that affects leak integrity exists at the gasketed joint between Hanford Connectors (HC) and nozzles. Review of the HCA Crane Log over a three-year period indicates the HCs periodically leak. Savannah River National Laboratory (SRNL) was requested to perform gasket compression tests that evaluate gasket relaxation characteristics, i.e., creep, and provide improvements to the HC installation process that can reduce or eliminate the frequency of leaks at HCs. A HC torque value of 312 ft-lbf is the target value used because it is the minimum torque value, with 95% confidence, when an unlubricated HC is impacted for standard sealing period of 9 seconds. In the HCA pipe-jumper connections are installed remotely by using HCs. HCs inherently add to the uncertainty of a gasketed joint because a calibrated torque value cannot be applied. The gasket material is a special material made of asbestos-fiber impregnated with Teflon. This material has been used for decades and generally performs well but does leak intermittently as gaskets creep and preloaded torque is lost. Gasket types are either snap-ring (SR) or lampshade (LS). Both gasket types cover the same seating area. The SR gasket is designed to remain with the jumper when the jumper is removed leaving the nozzle faces clean of any debris. SR gaskets are initially installed when the jumper is fabricated. Changing gaskets is time consuming and the radioactive environment presents risks to personnel. When jumpers are removed LS gaskets could be used to better seal the connection to reduce down time and ALARA concerns. That is, an LS gasket would be remotely positioned over the nozzle and the HC reinstalled. Testing also included two situations not typical of plant operation. One situation was the use of the non-standard double-layered gaskets (DG), and the other situation was the use of a LS without removal of the installed SR gasket. The DG was used when the supply of LS gaskets ended, so that testing could continue. It was of interest to demonstrate this gasket’s creep response because the DG, which was two gaskets glued together, was thicker than the LS gasket and the thicker material was expected to elicit a greater creep characteristics. The second situation of using two gaskets together by installing an LS gasket without removing the installed gasket is not allowed by the ASME Codes. However, by demonstrating the creep history of such a gasket arrangement it could be considered as an option that could possibly seal a connection while significantly reducing operational down time and personnel risk. Forty tests were conducted, and the overall results demonstrate the initial applied force and how the gasket changes with time. The change in sealing force, in terms of torque, was then measured to obtain a force history as the gaskets creep. This report documents the results of those tests as well as the method used to perform the tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Shear Assisted Processing and Extrusion (ShAPE) of Lightweight Automotive Components (CRADA 418)

Shear Assisted Processing and Extrusion (ShAPE) was developed to manufacture non-circular multicell profiles from secondary aluminum. This was accomplished by integrating a porthole die approach within the rotating ShAPE process. Complexity of the profile geometry was deliberately advanced over the course of the project from round to square, to asymmetric trapezoidal, to two-cell asymmetric trapezoidal. Aluminum alloy 6063 in the form of briquettes (compacted shreddings and engineered machining chips) and castings from pre-consumer industrial scrap were utilized as the feedstock. Tensile properties were shown to exceed the ASTM minimum standard and ASM typical values with the best results reaching yield strength = 247 ± 10 MPa, ultimate tensile strength = 271 ± 10 MPa, and uniform elongation = 16.5 ± 2.4%. These values were achieved for porthole extrusion using unhomogenized castings made from 100% aluminum 6063 industrial scrap. The ability to extrude unhomogenized billet was made possible by in situ conversion of plate-like β-type Fe-rich intermetallics to more extrudable needle-like α-type which is not possible with conventional extrusion. By eliminating the need to dilute iron with primary aluminum during the recycling process, savings of >90% on lifecycle carbon footprint and >50% on embedded energy could ultimately result in lower cost, more environmentally friendly, automotive components.

42 ENGINEERING↗

SAS4A/SASSYS-1 Commercial Grade Dedication Example Report for a Generic Sodium Pool-Type Fast Reactor Application

In the U.S., a key component of the commercialization of advanced reactors is completion of a license application, which must ultimately be approved by the Nuclear Regulatory Commission (NRC). The approval of the license application by the NRC is contingent on satisfactory demonstration of the design basis and the response of the advanced reactor design to transient and accident scenarios using accepted codes and methods. This report describes the qualification and dedication requirements that the advanced reactor safety analysis system software SAS4A/SASSYS-1 are expected to need to fulfill to be used for sodium-cooled pool-type fast reactor licensing. The qualification and dedication requirements are identified through performance critical characteristics and evaluation model acceptance criteria representative of the advanced reactor design considered for licensing. This document captures, additionally, the verification process developed to demonstrate that the software fulfills the qualification and dedication requirements for a generic sodium-cooled pool-type fast reactor as part of the commercial grade dedication process. Like most software that has primarily existed in the research and development space, the most significant challenge facing SAS4A/SASSYS-1 for use in a licensing framework is the availability of a documentation basis describing the code pedigree. SAS4A/SASSYS-1 has been used for licensing of the fast flux test facility (FFTF) and the JOYO sodium-cooled fast reactor in Japan, as well as the design of the CRBR Plant. However, the historical verification and validation (V&V) activities supporting SAS4A/SASSYS-1 development do not align with modern software quality assurance (SQA) and V&V requirements. Two approaches to use of SAS4A/SASSYS-1 in a commercial licensing framework have been identified: commercial-grade dedication (CGD) and software qualification. The methods and requirements prescribed in the ASME NQA-1-2008/2009 Standard and Regulatory Guide 1.203 on the evaluation model development and assessment process (EMDAP) have been used as guidance to define the CGD and qualification processes, respectively. A qualification and dedication requirements matrix has been developed which utilizes fundamental software verification. In this process, software verification is defined as a software quality process aimed at defining software requirement specifications, developing software design documentation, and performing and documenting acceptance testing of the code against requirements. A key element of software qualification and dedication includes determination of software acceptance with respect to critical characteristics relevant to the functional requirements of the software. To assist with identification of cross-cutting transient phenomena and functional requirements, domestic SFR vendor designs have been reviewed to identify a reference SFR design. For this report, the reference design is defined as a pool-type reactor with metal alloy fuel, a liquid-metal intermediate heat transport system, and passive decay heat rejection systems. Given this reference, a series of high-level cross-cutting phenomena was identified for a general class of single-fault undercooling or reactivity insertion transients that scopes the design basis space, with the goal of assisting with prioritization of documentation development efforts for key transient models in SAS: 1) Reactivity feedback response prior to scram; 2) System-wide thermal inertia; 3) Transition in natural circulation flow regime in heat removal systems; 4) Decay heat generation; 5) Steady-state fuel characterization; 5) Clad/fuel behavior at elevated temperatures; 6) Point kinetics and decay heat; 7) Pump coastdown behavior; 8) Core flow redistribution in loss of forced convection; 9) Pool stratification. As a demonstration of CGD of SAS4A/SASSYS-1 for a sodium pool reactor, a software qualification and dedication gap analysis as it relates to code documentation has been performed. This effort leverages the framework established as part of the SAS4A/SASSYS-1 SQA Program. This CGD demonstration provides a framework that vendors can build upon to demonstrate the applicability of the SAS4A/SASSYS-1 software for licensing a sodium-cooled pool-type fast reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Regulatory Treatment of Non-Core Sources of Radioactivity for Advanced Reactor Designs

The recent resurgence in advance (non-light water) reactor development has been paralleled by the development of risk-informed performance-based (RIPB) licensing pathways. Specifically, the creation of the RIPB Licensing Modernization Project (LMP) approach and subsequent endorsement by the U.S. Nuclear Regulatory Commission (NRC) now provides advanced reactor vendors with a defined RIPB method to develop an affirmative safety case for licensing. In addition, the Technology Inclusive Content of Applications Project (TICAP) has published guidance on developing a license application based on the LMP approach. To support the utilization of risk information as part of advanced reactor design and licensing efforts, the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Joint Committee on Nuclear Risk Management (JCNRM) has developed a probabilistic risk assessment (PRA) standard for advanced reactors. The standard, which was formerly approved by the American National Standards Institute (ANSI) in 2021 and recently endorsed by the NRC in trial use Regulatory Guide (RG) 1.247, is an integral standard, covering from initiating events to offsite consequence. A major feature of the standard is that it permits the inclusion of any source of radioactivity material at the site within the plant PRA. Therefore, non-core sources of radioactivity, such as fuel storage, fuel processing, and purification systems, can be included within a single comprehensive plant PRA. For those advanced reactor vendors utilizing a RIPB licensing approach, there is an opportunity to include the non-core sources of radioactivity within the RIPB framework for licensing decision-making, such as the categorization of events, classification of structures, systems, and components (SSCs), and evaluation of the adequacy of defense-in-depth (DID). For advanced reactor designs that contain multiple non-core sources of radioactivity, or for monolithic plant sites that include associated fuel facilities, this approach could potentially simplify licensing applications through the use of a single, uniform, and consistent decision-making framework across all radioactive sources at the site. In addition, a RIPB approach could provide additional insights regarding plant behavior, flexibility regarding licensing decision-making, and potentially allow the use of risk information as part of the plant oversight process. Risk-informing these aspects of advanced reactor licensing would also be consistent with the NRC’s risk policy statement. However, there is diverse regulation and guidance regarding the licensing of non-core sources of radioactivity and generally limited experience using RIPB approaches for the evaluation as part of licensing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Fundamental Graphite Irradiation Behavior Research at ORNL for FY23

This letter report is in submission of completion of the Level 3 milestone number within the larger Advanced Reactor Technologies Gas Cooled Reactor program at Oak Ridge National Laboratory (ORNL). The focus of this year’s efforts is the aggregation of background information about the irradiation experiments for property change in H-451 graphite, getting the irradiation-induced property change of IG-110 and ETU-10 written up in journal articles for publication, and involvement with ASTM D02.F0 Manufactured Carbon and Graphite Products and the ASME Working Group on Nonmetallic Design and Materials (SG HTR)(BPV III). This report will document the status of these activities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Additive Manufacturing Flaw Assessment Methodology

An evolution fatigue data and flaw tolerance of components produced using the Powder Bed Fusion (PBF) Additive Manufacturing (AM) process is documented in this report. Initial differences in fatigue data for AM components compared to smooth bar fatigue data indicated a need for a more detailed analysis of AM data available in technical literature. The investigation was initiated to support the development of a fatigue analysis methodology for AM components to support the of codification of AM technology for pressure equipment. The project was initiated to collect and analyze stainless steel 304L and 316L AM fatigue test data and corresponding process and quality information to develop S-N and E-N based fatigue data representation. Additional AM fatigue test data including Inconel Ti-6-4 and aluminum alloys were also considered for comparison purposes Metallic AM parts tend to contain various forms of defects distributed throughout the part. If an AM part is subjected to fatigue loading in service, a fatigue analysis needs to be performed during the design process to ensure an acceptable service life for the part. Post-process machining and polishing do not to improve fatigue resistance in any significant degree. The low cycle fatigue regime is of particular interest to this project in support of flaw acceptance criteria currently under development by ASME’s BPTCS/BNCS. Internal defects become exposed as external surface defects during machining for the machining of the AM part to final dimensions. This implies that as long as inherent AM defects are within a controlled limit in terms of both size and distribution characteristics, the corresponding fatigue test data in terms of either S-N (stress life) or E-N (strain life) can be investigated and characterized to establish fatigue properties of AM parts for design and fatigue evaluation purposes. The resulting S-N or E-N curves and their scatter bands can be used to derive fatigue design allowable stress values by capturing the effects of distributed discontinuities within an acceptable limit.

36 MATERIALS SCIENCE↗

Development of an accelerated creep testing technique via multi-specimen loading

This report summarizes a research effort undertaken by Argonne National Laboratory to accelerate the development of creep data package for code qualification. The goal is to explore new creep testing techniques to shorten the overall time needed to generate high-quality data to support an ASME code case on laser powder bed fusion 316H stainless steel. Progresses on the design and implementation of a 3-specimen load train and an innovative strain measurement method with direct potential current drop (DCPD) technique are reported.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Progress Report on Alloy 709 Base Metal Code Case Testing at ORNL in FY 2023

A collaborative research and development effort in support of the Alloy 709 Code Case qualification in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors is being carried out at Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL) and Argonne National Laboratory (Argonne). In FY 2023, ORNL has continued to conduct a subset of the Code Case testing for tensile, creep rupture, fatigue and creep-fatigue on Alloy 709. This report also updates the key Code Case testing status and results on the first two commercial heats and the preliminary results on the third commercial heat of Alloy 709. The three commercial heats of Alloy 709 are all in plate product form.

36 MATERIALS SCIENCE↗

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↗

Research Plan and Preliminary Results in Developing the Fabrication Parameters for Alloy 709 in Different Product Forms―Grain Coarsening Temperature Evaluation

The Advanced Reactor Technologies (ART) Program has established a multi-year plan to develop Alloy 709 advanced stainless steel (A709), generate the data package and develop material-specific design parameters in qualifying it as a new structural material for Class A component design in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors. In collaboration with material vendors, the Advanced Materials Development activities under ART have successfully scaled the A709 plate form production from a laboratory heat of 500 pounds to commercial heats totaling 133,000 pounds of A709 plate fabricated from three heats. The goal of the overall A709 development program is to establish the necessary microstructural and mechanical properties relationship for A709 to ultimately develop fabrication parameters for other product forms such as bars, piping and forging using the available ART A709 materials. The objective of this A709 development work at ORNL in FY2023 is to experimentally determine grain coarsening behavior for the A709 heats and to experimentally generate the continuous cooling precipitation (CCP) diagram for A709 using the as-rolled commercial heat plate materials. Integral to this work is the characterization of the as-rolled materials and the determination of an effective solution annealing process. This report summarizes the work performed to identify the grain coarsening temperature for commercial heat 58776-3RB fabricated by G. O. Carlson and heat 529900-02 fabricated by Allegheny Technologies Incorporated (ATI) Flat Rolled Products.

36 MATERIALS SCIENCE↗

FY23 Status Report for A709 Code Case Creep 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 American Society of Mechanical Engineers (ASME) Alloy 709 Code Case development. In Fiscal Year (FY) 23, 11 new tests were initiated and 11 tests were ruptured, some of those were initiated in FY21 or FY22. This report presents the creep data and the metallographic observations on selected rupture specimens.

36 MATERIALS SCIENCE↗

Measuring the Dissolution of Cr and Fe at 550°-750°C in FLiNaK and FLiBe

Assessing the compatibility of 316H stainless steel with molten fluoride salts has been identified as a key research topic for molten salt reactor development. Current developers are interested in 316H stainless steel due to its ready availability and ASME code qualification. Previous studies of 316H stainless steel compatibility in molten fluoride salts have shown general agreement that Cr is selectively removed from the alloy during the exposures in molten salts. However, the extent of Cr dissolution depends on the initial purity of the salt and Fe also readily dissolves in the salt. In order to model dissolution, it is necessary to understand how time and temperature affect the dissolution rates of the alloying elements which may lead to saturation of these elements in the salt. To study the dissolution of 316H, Cr and Fe specimens were exposed in FLiNaK and LiFBeF2 (FLiBe) salt at three different temperatures (550°-750°C) and three different durations (100-2000 h) in isothermal capsule experiments.

36 MATERIALS SCIENCE↗

Thermal Hydraulic Experimental Test Article - Fiscal Year 2023 (Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that is used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility simulates nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field may be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility was designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary coolant and secondary coolant system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. To date a test matrix has been completed utilizing the primary system of THETA. These tests, along with computational fluid dynamics and systems code models, determined the heat transfer across the core barrel and intermediate heat exchanger outlet was too great to effectively represent scaled thermal hydraulic phenomena of a liquid metal cooled reactor. Therefore, a significant effort was made to remove the primary system from the METL 28” test vessel #4 and clean the residual sodium from the primary system to facilitate upgrades. Thermal insulation was then incorporated in the core barrel and intermediate heat exchanger outlets. The primary system was then replaced, and a series of tests were performed to assess the performance of the thermal insulation. With primary system testing and upgrades complete, the secondary system could then be brought online. The tube side of the shell-and-tube intermediate heat exchanger was installed onto the primary system flange to begin installing the secondary system. The support structure for the secondary system was then erected on the METL mezzanine alongside the THETA primary system to facilitate installation of the secondary system components (sodium-to-air heat exchanger, flowmeter and pump). The piping and expansion tank were welded into the secondary system. Non-destructive examination of the secondary system welds was completed in order to satisfy ASME B31.3 pipe code for class M process fluids. The heating system and insulation were then added to prepare the system to be filled with sodium. The ancillary electrical equipment was installed which included the pump control box, blower VFD, pipe heater control system, etc. The secondary system will be filled, and a test matrix will be completed in early FY2024.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Integrated Issues and Risk Management: A Theoretical Framework Overview

The contractor requirements document for DOE O 226.1B, Implementation of Department of Energy Oversight Policy, requires DOE/NNSA contractors to establish an assurance system that includes, among other things, “Rigorous, risk-informed, and credible self-assessment and feedback and improvement activities. Assessment programs must be risk-informed, formally described and documented, and appropriately cover potentially high consequence activities” and “Contains an issues management process that is capable of categorizing the significance of findings based on risk and priority and other appropriate factors….” However, the term “risk-informed” is not defined in this or any other DOE order, and no formal guidance on how to integrate the two concepts currently exists. The Risk Management Guide for Defense Programs released by NA-18, Office of Systems Engineering and Integration (SE&I), states it is “a framework and general guidance to program office personnel on the effective management of program risks and issues”, however it then defines issues as “events with 100% likelihood of affecting program objectives” and states “unless specified otherwise, the term “risk” will also serve to represent issues for the remainder of this plan,” severally limiting its ability to provide adequate guidance on this topic. Outside of DOE scope, the U.S. Nuclear Regulatory Commission (U.S. NRC) imposes similar requirements. ASME NQA-1-2015 Requirement 16 states “Conditions adverse to quality shall be identified promptly and corrected as soon as practicable. In the case of a significant condition adverse to quality, the cause of the condition shall be determined, and corrective action taken to preclude recurrence. The identification, cause, and corrective action for significant conditions adverse to quality shall be documented and reported to appropriate levels of management. Completion of corrective actions shall be verified”. The purpose of this document is to provide a best-in-class framework for an integrated risk and issues management process. This process would provide a robust feedback loop between risk management and issues management to: Enhance risk identification and characterization, use risk handling principles to improve corrective action planning, and ensure regulatory compliance.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

Stress Relaxation Cracking of Alloys at Temperatures Higher Than 540°C

Type 347H stainless steel (347H SS), used in commercial concentrating solar power (CSP) thermal energy storage to store solar-salt at a temperature of 565°C, has been reported in the literature to be susceptible to stress-relaxation cracking (SRC). The welded heat-affected zone (HAZ) and fusion zone (FZ) of 347H SS, particularly in thick sections, are known to be susceptible to failure during post-weld heat treatment (reheat cracking). SRC could also occur after months or years under an elevated-temperature service environment. Two conditions must be present for failure to occur in the HAZ and/or FZ: 1) a modified or sensitized microstructure and 2) sufficiently high tensile residual stresses present at the elevated service temperature. The overarching goal of this project is to recommend SRC mitigation protocols to avoid susceptibility to fail through SRC at temperatures relevant for CSP. Post weld heat treatment conditions and alternative alloys are investigated as potential mitigation solutions to SRC. We used Gleeble thermomechanical simulation tests and finite element (FE) models to understand the susceptibility of 347H SS to SRC as a function of temperature, stress, and microstructure. We started with a literature review of weldability issues with 347H SS and techniques to mitigate SRC in 347H SS weldments. Next, we used Gleeble experiments to determine reheat cracking susceptibility in the simulated HAZ of 347SS weldments and an alternative alloy, 316L SS (NUCL 167 SPH) with boron added. We also performed Gleeble experiments to compare the reheat cracking susceptibility of 347H cross welded with two different fillers: E347, which is used in some commercial CSP TES tanks, and E16.8.2. We validated the experimental results with FE models of the residual stresses. We found that although the 316L (NUCL 167 SPH) is less susceptible to reheat cracking, it is slightly weaker than 347H and ASME BP&V codes limit its use to a service condition of 565°C. We also found that welds using E16.8.2 as the weld filler are less susceptible to failure than those using E347, likely due to the higher creep ductility of E16.8.2, and that it may be used as an alternative filler for repair welding of 347H welds or as the primary choice of filler for newly developed weld joints. We also found that post weld heat treatment could be a viable solution for mitigating stress in E347-347H SS thick, constrained welds, like those found in CSP tanks, and propose several options for mitigating SRC in existing and future TES tanks.

14 SOLAR ENERGY↗

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↗