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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.

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At least 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

STRESS CORROSION CRACKING OF ALLOY 152 WELD BUTTER NEAR THE LOW ALLOY STEEL INTERFACE

The objective of this work was to obtain SCC growth data in Alloy 152 weld butter near the interface with Low Alloy Steel (LAS), which is a region where some dilution of Cr was expected to have occurred, thus presumably exhibiting an increased SCC-susceptibility vs. the bulk of the weld. The LAS piece used in this application was Alloy 533-Gr B from the Midland reactor lower head, and the Alloy 152 weld butter received a prototypical Post Weld Heat Treatment (PWHT) prior to joining by Alloy 152 to an Alloy 690 piece according to a procedure qualified to ASME IX. The compact tension specimens for SCC testing were aligned in the first layer of the Alloy 152 butter. The experimental approach based on tracking environmental enhancement vs. location was successful in identifying SCC-susceptible locations, and SCC rates ranging from 10-12 m/s to as high as 10-10 m/s were measured. The post-test examination of the specimens found that the fracture had the intergranular/interdendritic appearance typical of welds, and that the propagation was arrested wherever an intersection with the LAS occurred. The large range of SCC rates measured does not appear to correlate well with the local concentration of Cr (approx. 25% at the SCC locations), and, in fact, low Cr (20%) – high Fe “streaks” seemed to slow/arrest crack propagation. In short, simple “Cr dilution” does not seem to fully account for the “SCC-susceptible” microstructure that yielded the 10-10 m/s growth rate in this weld.

Alloy 152↗
DOE OSTI2016

USE OF SMALL SPECIMENS FOR FRACTURE TOUGHNESS EVALUATION OF RPV STEELS

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of standard Charpy specimen or its broken half may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs and most likely to be used in advanced reactors as per ASME code.

Sokolov, Mikhail A↗
DOE OSTI2014

Albuquerque Regional Training: The Third Seminar on Surface Metrology for the Americas May 12-13 2014

The Third Seminar on Surface Metrology for the Americas (SSMA) took place in Albuquerque, New Mexico May 12-13, 2014. The conference was at the Marriott Hotel, in the heart of Albuquerque Uptown, within walking distance of many fantastic restaurants. Why surface metrology? Ask Professor Chris Brown of Worcester Polytechnic Institute (WPI), the chair of the first two SSMAs in 2011 and 2012 and the chair of the ASME B46 committee on classification and designation of surface qualities, and Professor Brown responds: “Because surfaces cover everything.”

42 ENGINEERING↗
DOE OSTI2018

THE ROLE OF MATERIAL MODELING ON STRAIN RANGE ESTIMATION FOR ELEVATED TEMPERATURE CYCLIC LIFE EVALUATION

High temperature nuclear reactors operating in the creep regime are designed to withstand numerous cyclic events. Current ASME code rules provide two basic paths for evaluating creep fatigue and ratcheting under these conditions; one based on full inelastic analysis intended to provide a representative stress and strain history and the other based on elastic material models with adjustments of varying complexity to account for inelastic stress and strain redistribution. More recent developments have used elastic-perfectly plastic analysis to bound the effects of cyclic service. However, these methods still rely on the separate evaluation of fatigue and creep damage utilizing a damage interaction diagram. There is a procedure under current development that uses creep-fatigue data from key feature test articles directly without the use of the damage interaction diagram. However, it requires a reasonable representation of the strain range in a structure as an input. This work develops a simplified procedure based on elastic perfectly-plasticity analysis that can be used to represent the strain range in a structure in the steady state under cyclic loading conditions.

Messner, Mark C.↗
DOE OSTI2018

N-BAR PROBLEMS AS APPROXIMATIONS TO THE BREE PROBLEM

The Bree solution to the problem of a ratcheting cylinder under constant pressure and cyclic thermal load is a fundamental result in nuclear engineering and widely used as the technical basis for the ASME Boiler and Pressure Vessel Code and other design methods. However, because the loading conditions in the Bree problem are difficult to achieve experimentally there have been relatively few works experimentally examining the problem and extending it to other relevant design situations, for example cladded components. In contrast, 2-bar problems are widely studied experimentally and are relatively easy to setup. These 2-bar problems are thought to be representative of Bree-type geometries, but a formal connection has not been demonstrated. This work formally establishes the connection between the Bree cylinder and an n-bar problem – a coupled bar experiment with, in general, more than two bars linked in parallel. The connection suggests that n-bar experiments using a fairly limited number of bars might be an experimentally-accessible setup that better represents ratcheting phenomenon in actual nuclear pressurized components. Such experiments could test surrogate cladded or multi-material components by using bars of different materials. Finally, this work suggests control schemes that yield optimally efficient n-bar experiments – experiments that best replicates a Bree cylinder with a limited number of bars.

Messner, Mark C.↗
DOE OSTI2019

Neutron Residual Stress Mapping of Repaired Spent Nuclear Fuel Welded Stainless-Steel Canisters

Stress corrosion cracking (SCC) caused by atmospheric deposition and deliquescence of aggressive compounds such as chloride-containing salts or oxidized sulfur species may pose a potential issue to the extended service life of spent nuclear fuel (SNF) storage canisters. When a chemically susceptible material is exposed to a hostile environment (e.g., coastal salt air) for a sufficient length of time, SCC may occur at locations on the canister which have undergone high tensile residual stress. These locations are often found in weld zones and the heat-affected zones (HAZ) of welded joints. Ideally, these cracks could be repaired by traditional fusion welding techniques, but the highly localized heat input of the repair welding may introduce additional high tensile residual stress and could increase the risk for future SCCs. This paper presents the results of neutron residual stress mapping of as-welded and repaired stainless steel specimens. The as-welded specimens were provided by Sandia National Laboratories from a canister mock-up manufactured using procedures similar to those used for actual canister production. ASME-qualified gas-tungsten arc welding (GTAW) was used to repair the specimens, and post-repair residual stresses were measured using neutron diffraction at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR). As-welded and repaired specimen residual stress distributions are evaluated and compared.

Chatzidakis, Stylianos↗
DOE OSTI2019

{ital n}-bar problems as approximations to the Bree problem

The Bree solution to the problem of a ratcheting cylinder under constant pressure and cyclic thermal load is a fundamental result in nuclear engineering and widely used as the technical basis for the ASME Boiler and Pressure Vessel Code and other design methods. However, because the loading conditions in the Bree problem are difficult to achieve experimentally there have been relatively few works experimentally examining the problem and extending it to other relevant design situations, for example cladded components. In contrast, 2-bar problems are widely studied experimentally and are relatively easy to setup. These 2-bar problems are thought to be representative of Bree-type geometries, but a formal connection has not been demonstrated. This work formally establishes the connection between the Bree cylinder and an n-bar problem a coupled bar experiment with, in general, more than two bars linked in parallel. The connection suggests that n-bar experiments using a fairly limited number of bars might be an experimentally-accessible setup that better represents ratcheting phenomenon in actual nuclear pressurized components. Such experiments could test surrogate cladded or multi-material components by using bars of different materials. Finally, this work suggests control schemes that yield optimally efficient n-bar experiments experiments that best replicates a Bree cylinder with a limited number of bars.

Messner, M. C.↗
DOE OSTI2019

Neutron Absorber Considerations for the DOE Standardized Canister Powerpoint

The DOE Standardized SNF Canister will allow a broad range of DOE-managed fuels to be loaded, stored, transported, and disposed of while avoiding the need to re-open canisters after being sealed. A key safety function of the DOE Standardized SNF Canister is to ensure the material maintains a subcritical configuration for storage, transportation, and disposal. Researchers designed a basket made of a material which should ensure subcriticality (e.g. by geometry control and inclusion of neutron absorber) and be structurally sound. The basket material should resist the corrosive effects of a 10,000 year period in a breached container and still provide adequate neutronic isolation or degrade in a form which prevents a critical configuration of fissile material. This report summarizes the materials evaluated to ensure subcriticality of the DOE Standardized SNF Canister for credible disposal scenarios. Recommendations for future work include performing criticality evaluations using borated stainless steel for the basket material in the DOE Standardized SNF Canister and determining the solubility on boron’s corrosion products. If researchers determine the properties of borated stainless steel are insufficient, ANA needs to undergo more corrosion testing and continue the process to become an ASME code case for welded construction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Application of Strain Gages in a Highly Advective Environment for Measurement of Fluid-Structure Interactions

A multi-year experimental program has been conducted to develop a program for applying strain gages to prototypic, plate-type fuel for high performance research reactors for use during flow testing. Plate strain measurements were intended to support the qualification of new fuel materials for use in high performance research reactors under the Materials Management and Minimization (MMM) Fuels Qualification Program (FQP). A survey of currently available strain gaging materials and processes was undertaken with a focus on applications towards thin, wide aspect ratio flow channels and fuel plates exposed to high velocity flows. Special consideration was taken to minimize the effect of the strain gage installations on the flow field. This test program has resulted in the selection of appropriate strain gage construction, adhesives, protective coatings, and assembly techniques to allow the real-time acquisition of plate strain during flow testing of the Generic Test Plate Assembly (GTPA) within the Oregon State University (OSU) Hydro-Mechanical Fuel Test Facility (HMFTF). All processes have been documented in procedures developed and utilized within the ASME NQA-1 compliant quality assurance program under which the HMFTF operates. Matrix testing has been completed using the results of this work with successful measurement of plate deformation due to hydraulic loading during flow testing.

Weiss, Aaron↗
DOE OSTI2019

EFCOG Best Practice #65 Engineering Calculations

The purpose of this paper is to present guidance for developing engineering calculations to assure that such technical documents meet the quality requirements of 10 CFR 830 and DOE O 414.1C. More specifically, this paper presents guidance related to the following general topics: Development; Design Input/Assumptions; Documentation/Content; Format; Rigor; Review/Verification; Approval; Revisions. Guidance for this paper was also taken from ASME NQA-1, Quality Assurance Requirements for Nuclear Facility Applications and its predecessor ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear Power Plants. These standards clearly point to what can be described as a graded approach. This guidance presented in this paper is intended to apply to safety related structures, systems and components (SSCs) whose reliable performance prevents or mitigates accidents that could cause undue risk to the health and safety of the public. Extension beyond this scope to non-safety related SSCs represents good engineering practice.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗