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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Management Academy LANL Business Systems: Property Management, Course #31036

Los Alamos National Laboratory (LANL) is responsible for the efficient economical management of all government property in its stewardship. This training explains the role LANL managers have in managing, controlling, and disposing of government property. The Laboratory's goal is good asset management. By properly managing property across the facility, Laboratory managers can help ASM improve government property utilization and extend asset life, while reducing asset-related operating costs and expenditures.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Report From BPTCS Project Team On Evaluation Of Additive Manufacturing For Pressure Retaining Equipment

ASME is evaluating the use of additive manufacturing (AM) for the construction of pressure equipment. The information in this report assesses available AM technologies for direct metal fabrication of pressure equipment. Background information is included in the report to provide context for those not experienced in AM technology. Only commercially available technologies for direct metal fabrication are addressed in the report because these AM methods are the only viable approaches for the construction of pressure equipment. Metal AM technologies can produce near-net shape parts by using multiple layers of material from a three dimensional (3D) design model of the geometry. Additive manufacturing of metal components was developed from polymer based rapid prototyping or 3D printing. At the current maturity level, AM application for pressure equipment has the potential to reduce delivery times and costs for complex shapes. AM will also lead to a reduction in the use of high cost materials, since parts can be created with corrosion resistant layers of high alloy material and structural layers of lower cost materials.

42 ENGINEERING↗
DOE OSTI2016

Nondestructive Examination Guidance for Dry Storage Casks

In this report, an assessment of NDE methods is performed for components of NUHOMS 80 and 102 dry storage system components in an effort to assist NRC staff with review of license renewal applications. The report considers concrete components associated with the horizontal storage modules (HSMs) as well as metal components in the HSMs. In addition, the report considers the dry shielded canister (DSC). Scope is limited to NDE methods that are considered most likely to be proposed by licensees. The document, ACI 349.3R, Evaluation of Existing Nuclear Safety-Related Concrete Structures, is used as the basis for the majority of the NDE methods summarized for inspecting HSM concrete components. Two other documents, ACI 228.2R, Nondestructive Test Methods for Evaluation of Concrete in Structures, and ORNL/TM-2007/191, Inspection of Nuclear Power Plant Structure--Overview of Methods and Related Application, supplement the list with additional technologies that are considered applicable. For the canister, the ASME B&PV Code is used as the basis for NDE methods considered, along with currently funded efforts through industry (Electric Power Research Institute [EPRI]) and the U.S. Department of Energy (DOE) to develop inspection technologies for canisters. The report provides a description of HSM and DSC components with a focus on those aspects of design considered relevant to inspection. This is followed by a brief description of other concrete structural components such as bridge decks, dams, and reactor containment structures in an effort to facilitate comparison between these structures and HSM concrete components and infer which NDE methods may work best for certain HSM concrete components based on experience with these other structures. Brief overviews of the NDE methods are provided with a focus on issues and influencing factors that may impact implementation or performance. An analysis is performed to determine which NDE methods are most applicable to specific components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Report on the completion of the procurement of the first heat of Alloy 709

This report provides details on the completion of the procurement of the first commercial-sized heat of Alloy 709. The report is a Level 3 deliverable in FY17 (M3AT- 17OR1602053), under the Work Package AT-17OR160205, “Advanced Alloy Development” performed by Oak Ridge National Laboratory, as part of Advanced Structural Materials Program for the Advanced Reactor Technologies (ART). This work project supports the fabrication scale up effort for Alloy 709 that was started in FY16. The effort culminated in the placement of a Purchase Order in August 2016 with a commercial vendor to melt an Alloy 709 heat using industrial melt practice. Four ingots, totaling about 45,000 lb, had been bottom-poured from the melt in September 2016. Two of the ingots were hot rolled into 1.2”x60”x155” and 1.1”x60”x100” plates using standard hot rolling process in FY17. Some small test pieces were cut from the asrolled plates and sent to Argonne National Laboratory (ANL) for archival. The plates were then heat treated and surface pickled by the vendor. The plates were subsequently delivered to ANL and Oak Ridge National Laboratory (ORNL). Properties and microstructure screening were performed on these plates upon delivery in February 2017 at ANL. Several samples were cut from the as-rolled and heattreated plates and were analyzed for their microstructures, hardness values, grain sizes, and room temperature tensile properties. The results indicate that the scaled-up heat of Alloy 709 fabricated using commercial practice exhibit tensile properties that exceed the minimum values specified in the ASME Code Case for commercial heat of NF709. These plates will be used to support base metal testing for the 650°C, 100,000-h Alloy 709 Code Case development, for fabrication of weldments, and for the NEUP projects.

36 MATERIALS SCIENCE↗
DOE OSTI2017

V27 Test Report.

The V27 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 third 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 si x 1.5 pound charges of Composition C - 4 (equivalent to 11.25 pounds TNT) distributed around the vessel in accordance with the User Design Specification. Four subsequent tests using less explosive evaluated the effects of slight variations in orientation of the charges . All vessel acceptance criteria were met.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2017

Readiness Review of BWXT for Fabrication of AGR 5/6/7 Compacts

In support of preparations for fabricating compacts for the Advanced Gas Reactor (AGR) fuel qualification irradiation experiments (AGR-5/6/7), Idaho National Laboratory (INL) conducted a readiness review of the BWX Technology (BWXT) procedures, processes, and equipment associated with compact fabrication activities at the BWXT Nuclear Operations Group (BWXT-NOG) facility outside Lynchburg, VirginiaVA. The readiness review used quality assurance requirements taken from the American Society of Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008/1a-2009) as a basis to assess readiness to start compact fabrication.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

FY17 Status Report on Testing Supporting the Inclusion of Grade 91 Steel as an Acceptable Material for Application of the EPP Methodology

This report summarizes the experiments performed in FY17 on Gr. 91 steels. The testing of Gr. 91 has technical significance because, currently, it is the only approved material for Class A construction that is strongly cyclic softening. Specific FY17 testing includes the following activities for Gr. 91 steel. First, two types of key feature testing have been initiated, including two-bar thermal ratcheting and Simplified Model Testing (SMT). The goal is to qualify the Elastic – Perfectly Plastic (EPP) design methodologies and to support incorporation of these rules for Gr. 91 into the ASME Division 5 Code. The preliminary SMT test results show that Gr. 91 is most damaging when tested with compression hold mode under the SMT creep fatigue testing condition. Two-bar thermal ratcheting test results at a temperature range between 350 to 650o C were compared with the EPP strain limits code case evaluation, and the results show that the EPP strain limits code case is conservative. The material information obtained from these key feature tests can also be used to verify its material model. Second, to provide experimental data in support of the viscoplastic material model development at Argonne National Laboratory, selective tests were performed to evaluate the effect of cyclic softening on strain rate sensitivity and creep rates. The results show the prior cyclic loading history decreases the strain rate sensitivity and increases creep rates. In addition, isothermal cyclic stress-strain curves were generated at six different temperatures, and a nonisothermal thermomechanical testing was also performed to provide data to calibrate the viscoplastic material model.

36 MATERIALS SCIENCE↗
DOE OSTI2017

High Efficiency and Low Cost Thermal Energy Storage System

BgtL, LLC (BgtL) is focused on developing and commercializing its proprietary compact technology for processes in the energy sector. One such application is a compact high efficiency Thermal Energy Storage (TES) system that utilizes the heat of fusion through phase change between solid and liquid to store and release energy at high temperatures and incorporate state-of-the-art insulation to minimize heat dissipation. BgtL’s TES system would greatly improve the economics of existing nuclear and coal-fired power plants by allowing the power plant to store energy when power prices are low and sell power into the grid when prices are high. Compared to existing battery storage technology, BgtL’s novel thermal energy storage solution can be significantly less costly to acquire and maintain, does not have any waste or environmental emissions, and does not deteriorate over time; it can keep constant efficiency and operates cleanly and safely. BgtL’s engineers are experienced in this field and are able to design and engineer such a system to a specific power plant’s requirements. BgtL also has a strong manufacturing partner to fabricate the system such that it qualifies for an ASME code stamp. BgtL’s vision is to be the leading provider of compact systems for various applications including energy storage. BgtL requests that all technical information about the TES designs be protected as proprietary information. To honor that request, only non-proprietay summaries are included in this report.

25 ENERGY STORAGE↗
DOE OSTI2017

Report on Understanding and Predicting Effects of Thermal Aging on Microstructure and Tensile Properties of Grade 91 Steel for Structural Components

This report provides an update on understanding and predicting the effects of long-term thermal aging on microstructure and tensile properties of G91 to corroborate the ASME Code rules in strength reduction due to elevated temperature service. The research is to support the design and long-term operation of G91 structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY17 (M2AT-17AN1602017), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Report on the FY18 Uniaxial Material Model Testing and Key Feature Test Articles Testing of Grade 91

This report summarizes the results from experiments performed in FY18 on Gr. 91 steel (heat 30176). The testing of Gr. 91 has technical significance because, currently, it is the only approved material for Class A construction in ASME Section III Division 5 that is strongly cyclic softening. The focus of this investigation is the effect of pre-cyclic loading history on the responses of Gr. 91 under key feature testing with Simplified Model Test (SMT) creep fatigue and two-bar thermal ratchet loading. For the pre-softened SMT, the SMT life reduction ratio was found to be approximately the same as the pre-softening fatigue damage ratio. The pre-softening also affected the ratchet behavior for the SMT specimens. Under the same two-bar thermal ratchet test conditions, the pre-softened specimens showed faster ratchet ting rates or larger accumulated strains than those without pre-softening history. In addition, results from previous uniaxial mechanical tests in support of the viscoplastic material model development for Gr. 91 are summarized, and the new results from thermomechanical testing at high temperature range are reported.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Status of Metallic Structural Materials for Molten Salt Reactors

In the US and around the world, significant interest surrounds development and deployment of nuclear systems using molten salts to transfer heat. Several configurations are being explored, including the use of fluoride salts either as the primary coolant of a reactor with solid fuel or with the fissile material dissolved in the salt. Molten Salt Reactors (MSRs) with fluoride salt typically have a thermal neutron spectrum. There are also fast-spectrum reactors that could use chloride salts to transfer heat. As a result of operation of the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory during the 1960s, a system with uranium fuel dissolved in a fluoride-salt medium is the most well-established technology. A nickel alloy with about 6%Cr and molybdenum as a solid solution strengthening agent was developed and deployed for all structural applications in this reactor. The alloy was commercialized as Hastelloy N. During these experiments and in post-decommissioning characterization of material behavior, it was determined that the most significant challenges for structural materials are embrittlement, from helium introduced by transmutation of Ni, and corrosion and grain boundary embrittlement from the fission product tellurium. The MSRE had an outlet temperature of approximately 650°C. The mechanical properties of Hastelloy N are not sufficient to support long-term operation of an MSR above this temperature. Qualification of a material for use in the ASME Boiler and Pressure Vessel Code, Section III, “Rules for Construction of Nuclear Facility Components,” Division 5, “High Temperature Reactors,” will facilitate licensing with the Nuclear Regulatory Commission. Hastelloy N has not been qualified for use in nuclear construction, and significant additional characterization would be required for Code qualification. Given that this alloy is susceptible to He embrittlement and has limited high-temperature strength, it is not recommended that Code qualification be pursued. Instead, it is recommended that a systematic development program be initiated to develop new nickel alloys that contain a fine, stable dispersion of intermetallic particles to trap He at the interface between the matrix and particle and with increased solid solution strengthening from addition of refractory elements. Extensive screening of attractive alloy compositions for elevated temperature strength, microstructural stability, weldability, and resistance to He embrittlement (characterized using ion implantation) will lead to an alloy down-selection for commercialization and Code qualification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

HPHT 17TR8 Test Articles: Post-Test Characterization Following Pressure Burst Testing (Final)

The objective of this study was to investigate whether F22 forging material ’s properties change near the failure location, in the event that two full-scale test pieces are intentionally taken to burst failure. 1 The two full-scale test pieces were created per high-pressure high-temperature guidelines in API 17TR8 and ASME Section VIII references in those guidelines.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Optimization of Advanced Steels for Cyclic Operation through an Integration of Material Testing, Modeling and Novel Component Test Validation

The efforts in this project were focused in three primary objectives, provided below: • To develop the needed microstructural processing and performance relationships and associated material models for specific constituents in fabricated weldments (such as the parent material, heat affected zone regions and weld metal), • Apply these metallurgical relationships through modeling of a welded pressure bearing power plant component subjected to cyclic operational conditions under both mechanical and thermal loading, and • Validate the model through novel structural feature and component tests. An attempt was made to optimize the heat treatment schedule to reduce or eliminate the presence of boron nitride (BN) in the matrix of the Grade 92 material. BN has been shown to be a primary nucleation source for creep damage. Proper evaluation of material behavior, and as assessed in this research, included the influence of geometric notches (such as those from poor fabrication or design) and metallurgical notches (such as weldments or poorly processed parent material). Better practice methodologies were implemented in the theoretical design of a high-temperature, critical component for cyclic operation. It was demonstrated that vessel testing can be reliably performed and provides a critical link between standard testing, structural assessment and validation. While the goal of an improved Grade 92 steel was not achieved in this project, there is considerable value in applying the documented testing and assessment methodologies to assess material and/or component behavior. This is especially true for materials and/or designs which are expected to be subjected to flexible operating modes. Design of a cycling state of the art thick wall steam header for long term operation with a state of the art ferritic steel was a task in this project. Implementation of that task showed how many factors including plant operating mode, creep strength, creep ductility, fatigue strength, and resistance to crack growth should be considered and evaluated to properly design high temperature components for long term service (200,000 operating hours or more) in a modern Fossil energy power plant. The current ASME non-nuclear boiler design code does not mandate this level of complex analysis nor does it provide all the metallurgical and mechanical properties data needed to perform this kind of design analysis.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2018

High Air-Flow Aerosol Filtration Testing

Pressure losses and aerosol collection efficiencies were measured for fibrous filter materials at air-flow rates consistent with high efficiency filtration (hundreds of cubic feet per minute). Microfiber filters coated with nanofibers were purchased and fabricated into test assemblies for a 12-inch duct system designed to mimic high efficiency filtration testing of commercial and industrial processes. Standards and specifications for high efficiency filtration were studied from a variety of institutions to assess protocols for design, testing, operations and maintenance, and quality assurance (e.g., DOE, ASHRAE, ASME). Three materials with varying Minimum Efficiency Reporting Values (MERV) were challenged with sodium chloride aerosol. Substantial filter loading was observed where aerosol collection efficiencies and pressure losses increased during experiments. Filter designs will be optimized and characterized in subsequent years of this study. Additional testing will be performed with higher hazard aerosols at Oak Ridge National Laboratory.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2018

Notch Effects on the Short Term Rupture Properties of Alloy 617 Base Metal and Weldments

Alloy 617 is currently undergoing approval for inclusion in the ASME Boiler and Pressure Vessel Code in Section III Division 5 for use in high temperature nuclear reactors. Prior to use in construction of these reactors, however, it must be approved by the Nuclear Regulatory Commission. Several concerns relating to the effects of geometric discontinuities, such as notches and multi-axial stress states (as opposed to uniaxial stress states typically present in laboratory run creep tests), must first be addressed. Alloy 617 has been found to be notch-strengthening in short term (1000-2500 hour) creep testing for both base metal and weld metal and will fail preferentially in a straight gauge section rather than at a sharp notch, provided that the applied stresses are similar in both the straight gauge and notch sections. Additionally, specimens designed with semi-circular notches (or U notches, as they are referred to in this report) to impose multi-axial stress states on the specimen during creep-testing were found to either maintain similar creep-rupture-lives as uniaxial stress specimens tested at similar conditions, or have creep-rupture-lives that were significantly longer. This work shows that notches and multi-axial stress states do not appear to raise concerns with relation to the use of design rules created for Alloy 617 using smooth, straight-gauge test specimens and uniaxial creep tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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

Evaluation of methods to determine strain ranges for use in SMT design curves

This report describes progress on developing a new method for creep-fatigue design that reduces over conservatism, improves the treatment of elastic follow up effects, and simplifies the design procedure, when compared to the existing ASME Boiler and Pressure Vessels Code, Section III, Division 5 method for creep-fatigue design. Specifically, this report describes the development of one of the three components of the new creep-fatigue design method: a design analysis method to represent the steady cyclic strain range in a component in elevated temperature service. The method developed here uses an elastic perfectly plastic (EPP) analysis with a pseudoyield stress selected to bound the width of the steady state stress-strain hysteresis loop. The report also provides a description of the in-development creep-fatigue design method, including a plan to complete the method by developing design curves and a method to account for stress multiaxiality.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Initial development and extension of EPP methods to Grade 91

This report describes modifications to the existing elastic perfectly plastic (EPP) analysis ASME Nuclear Code Cases N-861 and N-862 required to apply the Code Case design methods to Grade 91 steel. The critical difference between the existing Code Cases, applicable only to 304H and 316H stainless steel, is that they do not account for material cyclic softening. Grade 91 is a softening material at elevated temperature. As such, modifications to the existing Code Case rules are required before they can be conservatively applied to Grade 91.This report describes modified EPP methods applicable to softening materials like Grade 91. Additionally, the report describes the specific technical and administrative changes to the existing EPP Code Cases that will be required to support Grade 91.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗