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At least 361 records · Page 20

Bifurcation Buckling Analysis and Non-Linear Collapse Analysis of Teardrop Shaped Vacuum Chamber

The Spallation Neutron Source (SNS) accelerator is being upgraded to increase the beam power from 1.4MW at 1GeV to 2.8MW at 1.3GeV. The currents in the middle two injection chicane magnets cannot simply be scaled up to accommodate the increased injection energy of 1.3GeV due to potential excessive H- stripping; the magnets must be replaced with longer, lower-field magnets and the associated vacuum chambers need to be redesigned. A new teardrop-shaped vacuum chamber was initially designed to accommodate the new magnets and the updated beam paths and instrumentation. This paper focuses on the structural stability study of the teardrop shape vacuum chamber based on buckling analysis. Protection against collapse from buckling according to the ASME BPVC requirement has been evaluated in depth. First, a Type-1 bifurcation buckling analysis using a linear eigenvalue solution to determine the critical load factor was performed. Subsequently, a Type-3 nonlinear collapse analysis was conducted using the static Riks method with elastic-plastic material properties and imperfections explicitly considered in the model geometry. The critical buckling load for the teardrop shape vacuum chamber was confidently estimated based upon this two-stage approach.

Jiang, Hao↗

STRUCTURAL MODELING TO SUPPORT POST-YIELD ACCEPTANCE CRITERIA FOR SPENT NUCLEAR FUEL CLADDING

Spent nuclear fuel (SNF) is evaluated for structural failure during storage and transportation scenarios. The U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) program has sponsored significant research in quantifying mechanical loads on SNF during storage and transportation scenarios using experimental and modeling methods. The SFWST program has also performed significant research on measuring the mechanical behavior of irradiated SNF as defueled cladding segments and cladding with fuel pellets to measure composite behavior. This paper considers some of the key material data from the Sibling Pin testing and uses structural modeling and analysis methods that have been informed by testing to consider post-yield acceptance criteria for SNF cladding structural analysis. Test data published by Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL) are the foundation for informing the material behavior of the models developed in this study. In particular, four-point bend (4PB) tests of fueled and defueled cladding segments provide significant information about the bending failure mode of SNF. ORNL’s 4PB test data is on fueled cladding segments, so the composite behavior of SNF is demonstrated. This paper describes PNNL’s coincident beam model that was developed to approximate the composite behavior of SNF. This paper also presents PNNL’s structural dynamic finite element models of a cask tip-over scenario, which is predicted to cause the strongest mechanical loads on SNF of all postulated storage and transportation scenarios. SNF bending loads predicted in the cask tip-over scenario and cladding acceptance criteria beyond yield are considered, with justification based on the Sibling Pin test data. ASME Boiler and Pressure Vessel code stress intensity limits are also considered. The ultimate goal of this work is to aid in the justification of structural acceptance criteria for SNF cladding beyond the cladding’s irradiated yield strength for use in structural analysis of all storage and transportation scenarios.

Klymyshyn, Nicholas A.↗

Influence of Proton Activity Gaps between Electrodes on Open-Circuit Potential of H 2 /H 2 and H 2 /Air Cells

Polymer -electrolyte-fuel-cell open-circuit voltages (OCVs) are exactly defined by equation (1), where cathode and anode proton activities [(aH+)cathode and (aH+)anode, respectively] usually are identical, so the third term in the right-hand side of equation (2) is ignored. OCV=E0+RT/2F*ln(a1/2 O2*(a2 H+)cathode/aH2O)-RT/2F*ln((a2 H+)anode/aH2) (1) =E0+RT/2F*ln(a1/2 O2*aH2/aH2O)+RT/2F*ln((a2 H+)cathode/(a2 H+)anode) (2) Water vapor pressure is a colligative property that fundamentally correlates to electrolyte concentrations in aqueous solutions. Proton activity is a function of acid concentration, such as pH, when electrolytes are acids. In polymer-electrolyte membranes, water vapor pressure and acid concentration are understood as relative humidity (RH) and water uptake (λ), respectively, where λ represents number of water molecules per sulfonic acid molecule. Several investigations have reported the relationship between RH and λ, meaning that proton activities and associated water uptakes are intimately related to RH. In actual fuel cell operation, cathode RH is determined by ambient-atmosphere and/or humidifier RH(s), and anode RH depends on hydrogen-circulator RH. Therefore, RH is not always identical at both electrodes, and the difference between electrode RHs is considerable during dry operation of polymer electrolyte fuel cells. Therefore, the third term in the right-hand side of equation (2) may be significant for dry operation. We measured OCVs when hydrogen was supplied to both electrodes at 80°C. One electrode (A) was fixed at 30% RH, while RH at the other electrode (B) was varied (0, 5, 10, 20, and 30%). Measured OCVs varied from 0 to 75 mV. For fuel cell tests, electrode A was supplied with hydrogen at 30% RH; electrode B, oxygen at 0, 5, 10, 20, and 30% RH. OCVs deviated from that measured when RH at electrode B was 30%, increasing from 0 to 60 mV with decreasing RH at electrode B. Results are also shown in Figure 1. Proton activities of both electrodes were thermodynamically calculated. The Gibbs–Duhem relation was applied to obtain molar Gibbs free energies of water and sulfonic acid, and proton activity coefficient was calculated using the Gibbs free energy of sulfonic acid and the relationship between RH and λ1–4, assuming that protons and sulfonic anions show identical ionic-activity coefficients. OCVs were estimated using the third term in the right-hand side of equation (2). Results are shown in Figure 1. Fuel-cell current–voltage performance was poor when RHs at the anode and cathode were 30 and 20%, respectively. To determine kinetic current, we measured the oxygen-reduction reaction (ORR) using a rotating-disk electrode (RDE) in concentrated-acid aqueous solutions, which modeled catalyst-layer ionomers. Kinetic currents decreased with acid concentrations. References T. A. Zawodzinski, Jr., C. Derouin, S. Radzinski, R. J. Sherman, V. T. Smith, T. E. Springer and S. Gottesfeld , J. Electrochem. Soc., 140,1041 (1993) P. K. Das and A. Z. Weber, Proceedings of the ASME 2013 11th Fuel Cell Science, Engineering and Technology Conference, Fuel Cell 18010 (2013) V. A. Sethuraman, J. W. Weidner, A. T. Haug, S. Motupally,b and L. V. Protsailo, J. Electrochem. Soc., 155, B50 (2008) A. Kusoglu and A. Z. Weber, Chem. Rev., 117, 987 (2017) Figure 1

Yoshida, Toshihiko↗

BlendPATH (Blending Pipeline Analysis Tool for Hydrogen) [SWR-24-10]

BlendPATH provides case-by-case techno-economic analysis for potential projects where hydrogen is blended into a natural gas transmission pipeline. BlendPATH estimates 1) the transmission pipeline modifications and operating conditions necessary to blend hydrogen to a user specified volume faction of hydrogen in pipeline gas and 2) the incremental capital and operating expenses to prepare said transmission pipeline for hydrogen blending. BlendPATH is developed in Python and requires SAInt, a underlying commercial natural gas pipeline network modeling software, to run. Pipeline modification and operation condition estimation is guided by ASME B31.12. The intent of this software is to target application for projects in the initial project assessment stage and provide the user with the capability of assess promising opportunities before the use proceeds with further detailed pipeline evaluation based on a probable economic outcome.

Kee, Jamie↗

Compilation of Published Test Data for Cylindrical and Spherical Vessels under External Pressure

This dataset is a compilation of the results of previously-published failure tests of cylindrical and spherical pressure vessels, pipes, shells, and caps under external pressure. This data has been used for the article 'Simplified Formulas for External Pressure Design' by W. Kirkland and C. Bett, presented at the ASME Pressure Vessels and Piping Conference 2022. Paper number PVP2022-78354.

42 ENGINEERING↗

Optimization of a cyclone using MFIX and Nodeworks

Video depicting the optimization process of a cyclone on NETL's chemical looping reactor (CLR) using MFIX and Nodeworks. MFIX is used to model the cyclone using PIC. Nodeworks is then used to generate proposed geometry changes using a Latin hypercube. Each design is simulated, with an objective value being computed based on the cyclone efficiency and pressure drop. A Gaussian Process surrogate model is then constructed from the objective values. This surrogate model is then used by a differential evolution optimization algorithm to identify the optimal cyclone design. Details published here: Weber, J., Fullmer, W., Gel, A., and Musser, J. (February 4, 2020). "Optimization of a Cyclone Using Multiphase Flow Computational Fluid Dynamics." ASME. J. Fluids Eng. March 2020; 142(3): 031111. https://doi.org/10.1115/1.4045952 OSTI: https://www.osti.gov/pages/servlets/purl/1763893

cyclone↗

Pressure Spike in The LBNF Absorber Core s Gun Drilled Cooling Channel from an Accident Beam Pulse

The LBNF Absorber consists of thirteen 6061-T6 aluminum core blocks. The core blocks are water cooled with de-ionized (DI) water which becomes radioactive during beam operations. The cooling water flows through gun-drilled channels in the core blocks. The cooling water is supplied by the LBNF Absorber Radioactive Water (RAW) cooling system which is designed as per ASME B31.3 Normal Fluid Service [1]. An uninhibited beam accident pulse striking the water channels was identified as a credible accident scenario. In this study, it is assumed that the beam pulse hits the Absorber directly without interacting with any of the other upstream beamline components. The beam parameters used for the LBNF beam are 120 GeV, 2.4 MW with a 1.2 s cycle time. The accident pulse lasts for 10 μs. The maximum energy is deposited in the 3rd aluminum core block. For the sake of simplicity, it is assumed that the accident pulse strikes the 1 in. ID water channel directly. The analysis here simulates the pressure rise in the water during and after the beam pulse and its effects on the aluminum piping components that deliver water to the core blocks. The weld strengths as determined by the Load and Resistance Factor Design (LRDF) and the Allowable Strength Design (ASD) are compared to the forces generated in the weld owing to the pressure spike. A transient structural analysis was used to determine the equivalent membrane, peak, and bending stresses and they were com-pared to allowable limits.

43 PARTICLE ACCELERATORS↗

Optimization of a Welding Procedure for Making Critical Aluminum Welds on the LBNF Absorber Core Block

The LBNF Absorber consists of thirteen 6061-T6 aluminum core blocks. The core blocks are water cooled with de-ionized (DI) water which becomes radioactive during beam operations. The cooling water flows through gun-drilled channels in the core blocks. A weld quality optimization was performed to produce National Aeronautical Standard 1514 Class I quality welds on the aluminum core blocks. This was not successful in all cases. An existing Gas Tungsten Arc Welding Procedure Specification was fine tuned to minimize, in most cases, and eliminate detect-able tungsten inclusions in the welds. All the weld coupons, however passed welding inspection as per the piping code: ASME B31.3 Normal Fluid Service. Tungsten electrode diameter, type, and manufacturer were varied. Some of the samples were pre-heated and others were not. It was observed that larger diameter electrodes, 5/32 in., with pre-heated joints resulted in welds with the least number of tungsten inclusions. It is hypothesized that thinner electrodes breakdown easily and get lodged into the weld pool during the welding process. This breakdown is further enhanced by the large temperature differential between the un-preheated sample and the hot electrode.

43 PARTICLE ACCELERATORS↗

Design Guidance for High Temperature Concentrating Solar Power Components

This report provides guidance for the design of components for concentrating power facilities operating at high temperatures and undergoing high, secondary thermal stresses relative to the applied primary pressure stress. The design rules were developed for the design of Generation 3 CSP tubular receivers manufactured from Alloy 740H, but are generally applicable to a wide range of component types undergoing similar loads. Part 1 of the report provides procedural design rules for components to be used in conjunction with the 2019 edition of the ASME Boiler & Pressure Vessel Code. Part 2 provides corresponding design data for Alloy 740H. Part 3 of the report is a commentary describing the rationale behind the design rules and the data underlying the design material properties. Finally, Part 4 provides an extensive set of worked sample problems detailing the application of the rules to CSP components.

24 POWER TRANSMISSION AND DISTRIBUTION↗

V28 Test Report

The V28 containment vessel was procured by the US Army Recovered Chemical Material Directorate (RCMD) as a replacement vessel for use on the P2 Explosive Destruction Systems. It is the fourth EDS vessel to be fabricated under Code Case 2564 of the 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 nine (9) pounds TNT- equivalent for up to 637 detonations. This report documents the results of explosive tests that were done on the vessel at Sandia National Laboratories in Albuquerque New Mexico to qualify the vessel for explosive use. The primary qualification test consisted of six, 1.5 pounds charges of Composition C-4 (equivalent to 11.25 pounds TNT) distributed around the vessel in accordance with the User Design Specification. This test was repeated due to a lack of proper clamp settings. Two additional tests using less explosive were performed, one identical in configuration to a test performed in the V27 qualification series as a baseline for comparison, and one where the separation distance of the charges was increased to extend the V27 analysis of distributed load effects on the P2 vessel. All vessel acceptance criteria were met.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

Assessment of the Material Performance for TCR Applications (Progress Report)

The objective of Argonne’s materials research activities for the Transformational Challenge Reactor (TCR) program is to improve the understanding of material properties from additive manufacturing with the focus on understanding the creep and fatigue properties of additively manufactured materials. The material work conducted by Argonne provides support in developing and qualifying advanced materials and manufacturing processes to allow for innovative reactor design and licensing for the TCR. In FY 2020, Oak Ridge National Laboratory (ORNL) provided six ASTM standard sized creep specimens of additively-manufactured 316L stainless steel (AM 316L SS) for evaluation of the effect of post-build heat treatment on the creep property. Two different teat treatments were conducted on the as-built specimens, namely 1050°C/1h and 650°C/1h before creep testing. All specimens were creep tested at 550°C and 275 MPa to understand the effect of heat treatments. We have also received AM 316L SS printed in various forms including rods and sheets, wrought 316 L SS (as a reference material), and AM IN718 for tensile, creep and fatigue testing. Both ASTM standard sized specimens and subsized specimens are being fabricated or tested to contribute to a comprehensive mechanical property database of AM materials and to provide location-specific property data to the TCR digital platform. Mechanical testing at Argonne is conducted to ASME NQA-1 or its lab equivalent for quality assurance. This report presents the results of the creep tests at 550°C and microstructural characterization for the assessment of the effect of the post-build heat treatment on the creep behavior of AM 316L SS. It also discusses ongoing efforts on obtaining location-dependent properties of AM 316L SS.

36 MATERIALS SCIENCE↗

Initial development of a high temperature life prediction method directly accounting for variability in material properties

This report presents the initial development of a life prediction method that accounts for the variability of the material properties in Grade 91 steel. We account for material variability by both fitting a variable 3-parameter Weibull distribution to the rupture data of the material and by quantifying the effect that the creep deformation properties have on stresses via a Monte Carlo approach. To ensure a reasonable time frame when applying Monte Carlo to the creep stress, we propose a methodology that models the material as an extremely viscous Stokes flow with a non-Newtonian viscosity, therefore solving the problem in a single step. We use the final probabilistic model for structural failure developed in this document to evaluate a Flat Head Vessel designed with the Section III, Division 5 ASME Boiler Pressure and Vessel Code rules with a 100,000h design life and assess its probability of premature failure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Initial development for the introduction of provisions for irradiation effects in Division 5 rules

Heat pipe microreactor concept involves the use of a solid core block to encapsulate the nuclear fuel and the passive extraction of heat from the core block to the heat exchanger block by advanced heat pipes. The core block is a key component of the heat pipe microreactor design. It integrates the functions of reactor vessel, core structural components and fuel cladding of traditional Gen IV advanced reactor designs into one single structure. The effects of irradiation on the structural failure modes are outside the scope of the rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B for metallic Class A components. Methods to account for the irradiation effect were investigated in order to develop design methods that can be applied to the core block design in the heat pipe microreactor. This report presents an initial development of an approach to account for the irradiation effects in the design rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Thermal Hydraulic Experimental Test Article (FY2020 Status Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be 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 will simulate 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 will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being 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 and secondary 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. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2. The design work for the facility has been completed. Since the FY19 THETA status report, the THETA primary system components have been assembled and tested using de-ionized water. Some results of this testing at select Richardson and Reynolds numbers are included herein. The custom THETA sodium flowmeter (permanent magnet submersible sodium flowmeter) has been built and calibrated using sodium in an auxiliary facility against a NIST traceable vortex shedder type flowmeter. The permanent magnet flowmeter showed excellent performance, following the theoretical equation for measured voltage quite well, with an error < 3.5% at flow rates ranging from 3-24 GPM at temperatures up to 400 °C. The secondary sodium system for heat rejection from intermediate heat exchanger has been analyzed according to ASME code. All secondary side piping, fittings and a U-stamped air-to-sodium heat exchanger have been procured.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Progress Report on the Assessment of the Material Performance for TCR Applications

The objective of Argonne’s materials research activities for the Transformational Challenge Reactor (TCR) program is to improve the understanding of material properties from additive manufacturing with the focus on understanding the creep and fatigue properties of additively manufactured materials. The material work conducted by Argonne provides support in developing and qualifying advanced materials and manufacturing processes to allow for innovative reactor design and licensing for the TCR. In this report, we present the creep data at temperatures of 550, 600 and 650°C and stresses between 175 and 300 MPa for 316L stainless steel produced by a laser powder bed fusion process. We conducted five different post-build heat treatments, namely 650°C/1h, 750°C/1h, 800°C/1h, 900°C/1h and 1050°C/1h. Creep tests of the heat-treated specimens were performed under the same test condition, 550°C/275 MPa to evaluate and understand the effect of post-build heat treatment on the creep behavior of additively-manufactured 316L SS. Creep tests were conducted to ASME NQA-1 or its lab equivalent for quality assurance.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Evaluation of ASCE 4-16 and AISC 43-18 (Draft) for use in the Risk-Informed Performance-Based Seismic Design of Nuclear Power Plant Structures, Systems, and Components

This report describes the assessment of ASCE Standards 4-16 and 43-18 (Draft) for use in the Risk-Informed Performance-Based (RIPB) seismic design of structures, systems, and components (SSCs) at nuclear power plants. This work was performed for the U.S. Nuclear Regulatory Commission (NRC) Office of Regulatory Research (RES), to support potential endorsement of these industry standards for the design of nuclear power plants based on the RIPB approach. Currently, the NRC endorses a deterministic design approach for demonstrating the design adequacy of SSCs based on the Standard Review Plan (NUREG-0800) and NRC Regulatory Guides. In the RIPB approach, the design criteria are developed to achieve a target performance goal, which is defined by the annual frequency of occurrence of the design basis earthquake (i.e., Seismic Design Category) and the acceptable level of structural performance (i.e., Limit State) for the SSCs. ASCE 4-16 provides methods for performing a seismic analysis of structures to obtain the seismic response of these structures (e.g., building displacements, accelerations, in-structure response spectra) which are used in the design of the SSCs. It also provides methods for performing seismic analysis of SSCs to determine the seismic demands (e.g. member forces and displacements) needed to design individual SSCs. ASCE 43-18 (Draft) provides the criteria for the seismic design of SSCs using the seismic demands developed in ASCE 4-16. The use of ASCE Standard 43-18 (Draft), along with ASCE 4-16, provides the criteria for the seismic design of the SSCs. ASCE 43-18 (DRAFT) relies on other consensus codes and standards such as ACI 349 for reinforced concrete, AISC/N690 for steel structures, ASME Section III for pressure-retaining mechanical components and Containments, and IEEE-344 for Class 1E equipment. The goal of this technical review was to assess the adequacy of the provisions in these standards for use by the NRC in developing regulatory guidance for design of SSCs in nuclear power plants, based on the RIPB approach. The research reported herein describes the basis for acceptance of the new standards and identifies areas where additional staff guidance is needed for the seismic design of SSCs at nuclear power plants. This technical review has determined that ASCE 4-16 and ASCE 43-18 (Draft) provide an appropriate framework for the seismic design of SSCs at nuclear power plants using a Risk-Informed Performance-Based approach. However, some of the criteria therein warrant exceptions, qualifications, and/or clarifications.

42 ENGINEERING↗

Report on FY 2020 testing of Alloy 709 in support of EPP analysis

Research activities in support of the development of a 500,000 hr, 760ºC ASME Code Case (CC) for Alloy 709 are ongoing at Oak Ridge National Laboratory (ORNL), Argonne National Laboratory, and Idaho National Laboratory. Integration of the simplified Elastic-Perfectly Plastic (EPP) methods for the design of Alloy 709 reactor components is a requirement for the Alloy 709 CC. This report documents the research plans and initiation of testing in support of integrating EPP analysis methods for Alloy 709 at ORNL in FY 2020. Tests are designed and initiated to evaluate the effect of cyclic loading on the creep properties of Alloy 709. The preliminary results show that cyclic loading has an insignificant effect on the minimum creep rate, although additional tests are recommended to confirm this conclusion. A preliminary testing plan for multi-axial creep rupture has been established and the specimens are machined. Execution of the testing plan will continue in FY 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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↗