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At least 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Hanford Single-Shell Tank Leak Causes and Locations - 241-BY and 241-TY Farm

This document identifies 241-BY Tank Farm (BY Farm) and 241-TY Tank Farm (TY Farm) leak causes and locations for the 100-series leaking tanks in BY and TY Farm. The leak causes and locations report for all of the 100-series single-shell leaking tanks is one of the targets (M-045-91-T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F. The T04 target requires that the Department of Energy provide to State of Washington, Department of Ecology (Ecology) a report on the 100-series single-shell tanks which have been or will be identified as having leaked in RPP-32681, Rev. 0, Process to Assess Tank Farm Leaks in Support of Retrieval and Closure Planning, leak assessment reports. The leak assessment report for BY Farm, RPP-RPT-43704, Rev. 0A, Hanford BY-Farm Leak Assessments Report, lists one 100-series tank that either continues to be or is recommended to be classified as a leaking tank in BY Farm, 241-BY-103 (BY-103). The leak assessment report for TY Farm, RPP-RPT-42296, Rev, 0, Hanford TY-Farm Leak Assessments Report, lists four 100-series tanks that either continue to be or are recommended to be classified leaking tanks in TY Farm, 241-TY-103 (TY-103), 241-TY-104 (TY-104), 241-TY-105 (TY-105), and 241-TY-106 (TY-106). All of the other eleven 100-series tanks in BY Farm and two 100-series tanks in TY Farm are classified as “sound” or are identified in RPP-RPT-43704, Rev. 0A, and RPP-RPT-42296, Rev. 0, respectively, as requiring re-assessment of their classification per TFC-ENG-CHEM-D-42, Tank Leak Assessment Process. The TFC-ENG-CHEM-D-42 assessments are not part of the M-045-91-T04 target. This BY and TY Farm leak causes and locations document is part of a series of tank farm reports that identify leak causes and locations for 100-series leaking tanks. A summary and conclusions document will be issued, RPP-RPT-54909, Hanford Single-Shell Tank Leak Causes and Locations – Summary, that compiles the results from all of the leak causes and locations tank farm reports when they have been issued which will fulfill the T04 target requirements. The identification of four of the five leaking BY and TY Farm tank leak locations focused on the first indication of radiation detected in drywells as well as liquid level decreases as appropriate, with the exception of tank TY-104 which could not be identified. Leak detection laterals were not installed underneath the BY and TY Farm tanks. For TY Farm, direct pushes were installed in 2005 to further characterize the tank leaks which were used to collaborate with the drywell data. The BY and TY Farm leaking tanks were likely due to chemistry-corrosion from the storage of Tri-butyl phosphate (TBP) waste which is conducive to stress corrosion cracking and pitting. Thermal shock (rate of temperature rise) and tank construction conditions (cold weather ductile-to-brittle transition temperature impact effects) were also likely to have affected tank integrity. There appears to be very little contribution from tank design and liner bulging. However, some or all of the factors can act serially or together to contribute to tank liner failure. The leaking BY and TY Farm tanks all stored undiluted TBP waste whereas the sound tanks stored diluted TBP waste as well as other waste types.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Summary of available data for estimating chloride-induced SCC crack growth rates for 304/316 stainless steel

The majority of existing dry storage systems used for spent nuclear fuel (SNF) consist of a welded 304 stainless steel container placed within a passively-ventilated concrete or steel overpack. More recently fielded systems are constructed with dual certified 304/304L and in some cases, 316 or 316L. In service, atmospheric salts, a portion of which will be chloride bearing, will be deposited on the surface of these containers. Initially, the stainless steel canister surface temperatures will be high (exceeding the boiling point of water in many cases) due to decay heat from the SNF. As the SNF cools over time, the container surface will also cool, and deposited salts will deliquesce to form potentially corrosive chloride-rich brines. Because austenitic stainless steels are prone to chloride-induced stress corrosion cracking (CISCC), the concern has been raised that SCC may significantly impact long-term canister performance. While the susceptibility of austenitic stainless steels to CISCC in the general sense is well known, the behavior of SCC cracks (i.e., initiation and propagation behavior) under the aforementioned atmospheric conditions is poorly understood. A literature survey has been performed to identify SCC crack growth rate (CGR) studies conducted utilizing conditions that may be relevant to existing SNF interim storage canisters, the results of which are presented in this document. The data presented here have been restricted to those representing atmospheric corrosion of stainless steels due to deliquescence of marine salts, or marine salt components, on the metal surface. A suite of experimental studies representing both long-term field tests and accelerated laboratory tests has been identified. Potentially relevant data are summarized in Figures 1-1 (304 SS) and Figure 1-2 (316 SS). In the Figures, when a particular reference utilized a series of samples, the range is shown as a bar, and the average value shown with a symbol. A summary of the test methods, sample geometry, and environmental conditions for each study is given in Table 1-1. While the surveyed studies all explore SCC of austenitic stainless steels under atmospheric conditions, the methods through which each researcher approached the problem do differ, as illustrated in Table 1-1. The surveyed studies utilized a variety of metal treatments including as-fabricated, solution annealed, welded, and sensitized material. Furthermore, different surface treatments (polished vs ground) were also used. In addition, most of these studies were accomplished using techniques that are not generally accepted for high-fidelity crack growth rate measurements, and in cases where more traditional approaches were taken, these methodologies may not be applicable to the atmospheric conditions of interest here. The wide variety of methods and materials results in the observed large scatter in measured CGRs. Each of the data sets in Figures 1-1 and 1-2 is described in more detail in the following sections. A short summary of crack growth rates based on operational experience is also presented.

36 MATERIALS SCIENCE↗
NASA NTRS1995

Studies of localized corrosion in welded aluminum alloys by the scanning reference electrode technique

Localized corrosion in welded samples of 2219-T87 Al alloy (2319 filler), 2090 Al-Li alloy (4043 and 2319 fillers), and 2195 Al-Li alloy (4043 and 2319 fillers) has been investigated using the relatively new scanning reference electrode technique. The weld beads are cathodic in all cases, leading to reduced anode/cathode ratios. A reduction in anode/cathode ratio leads to an increase in the corrosion rates of the welded metals, in agreement with results obtained in previous electrochemical and stress corrosion studies involving the overall corrosion rates of welded samples. The cathodic weld beads are bordered on both sides by strong anodic regions, with high propensity for corrosion.

Danford, M. D.↗
DOE OSTI2017

FY16 Status Report: SNF Interim Storage Canister Corrosion and Surface Environment Investigations

This progress report describes work done in FY15 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

3013 Inner Container Closure Weld Region (ICCWR) Characterization by Wide Area 3D Measurement System (WAMS) Analysis (FY21 Progress Report)

As part of the 3013 Surveillance Program, through-wall penetration from stress corrosion cracking (SCC) of the 3013 inner containers has been identified as the most credible condition for failure withing the 50- years lifetime. Chlorides contained in Pu-bearing material, together with intra-canister humidity levels, metallurgical conditions, and internal stresses have been found to produce corrosion in the Inner Container Closure Weld Region (ICCWR) of the 3013 canister system. A Laser Confocal Microscope (LCM) is used as part of the 3013 Surveillance Program protocol to identify the prevalence of corrosion and corrosionrelated cracking in the ICCWR2. With the LCM, a close visual examination is made of the ICCWR surface along with measurements of corrosion-related features. LCM inspections produce immense amounts of image data that is time intensive to analyze. There is also a 9-year backlog of images, with approximately 49 canisters that must be evaluated. To expedite data analysis and reduce the amount of generated data, a Wide Area 3D Measurement System (WAMS) microscope has been added to the examination protocol. Although WAMS is a lower resolution microscope, small features of interest can be still identified. The advantage of collecting data for the full circumference using the WAMS is that it can take about 1/16 of the time needed with the LCM. Both systems offer capabilities that combined can be utilized to expedite the examination of the ICCWR. The WAMS is an efficient system for screening and identification of corrosion features while the LCM can be utilized to obtain higher resolution images areas identified by the WAMS. This report explains and justifies the data collection methods used with the WAMS. It includes a summary of the data generated by WAMS in FY21. The goals set for data collection in FY21 were met. A total of twenty-two DE’s were imaged, and analysis was carried out on eleven samples. The analyzed samples showed large numbers of potential cracks and pits distributed throughout the surfaces. Lastly, to explain the advantages of the WAMS, its capabilities were compared to those of a simpler microscope. Micrographs of the samples imaged in FY21 and those analyzed are included in the appendices.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

FY23 Update: Aerosol Sampling for the Canister Deposition Field Demonstration

This report describes the results of preliminary testing of aerosol monitoring equipment that will be used to continuously monitor the aerosol source term for the multi-year Canister Deposition Field Demonstration (CDFD). These data are required inputs for the development and validation of models for the deposition of dust and potentially corrosive salts on the surface of spent nuclear fuel (SNF) dry storage canisters. Surface salt loads correlate with the extent of corrosion damage on a metal surface, and potentially to the likelihood and timing of initiation of stress corrosion cracks. Aerosols will be monitored at the CDFD site using three instruments. A Dekati ® ELPI+ cascade impactor will be used for real-time monitoring of aerosol particle sizes. It will also collect dust in 14 size bins on impactor targets that can be chemically analyzed to determine the soluble salts present as a function of particle size. However, this instrument can only measure dried aerosols, with a diameter of <10 µm. The second instrument is a Topas laser particle size spectrometer, which provides real-time monitoring of aerosol particle sizes up to ~40 µm in size. It monitors both the ambient (potentially deliquesced) aerosol particle size distributions required for the dust deposition models and the distributions of the equivalent dried particles, allowing correlation with the Dekati ® data. However, it does not discriminate between inert dust particles and salt aerosols, and it does not retain samples of the different particle sizes for later analysis. The third instrument that will monitor aerosols at the CDFD site is a Clean Air Status and Trends Network (CASTNET) tower, which uses a multiple canister system to collect weekly samples for analysis to total suspended aerosol particle compositions and atmospheric gas concentrations. This status report describes work in FY23 to develop the capabilities for using these tools. In two training exercises, the cascade impactor and laser particle sizer were deployed in two different testing environments, one indoor and one outdoor. For the cascade impactor, the tests provided opportunities for the operators to familiarize themselves with impactor substrate preparation, and post-test sample removal and analysis. For the laser particle sizer, the tests were used to evaluate different instrument parameters, to determine the most appropriate settings for capturing transient events. Data and samples were collected for weeks to months for each test, and the results are presented here. In addition to the preliminary testing, contracts were developed with WSP Analytical Labs for sample preparation and analysis of the cascade impactor samples. The impactor tower from outdoor test was delivered to WSP and used to train the staff there in disassembly, sample extraction, sample analysis, and tower reassembly with new target substrates. These are tasks that WSP will be performing routinely for the CDFD project. The CASTNET system cannot be purchased or tested until an actual site has been selected for the CDFD test. Work for this FY has been restricted to preparation of contracts for purchasing the CASTNET tower, and for sample analysis, once the tower is in operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Mechanical response and microstructure evolution from multiaxial stress relaxation in textured zircaloy nuclear cladding

Modern energy markets provide motivation for nuclear reactors to increase operational flexibility and load following capacity of existing reactor designs. The operational power changes result in fuel pellet expansion and contraction resulting in pellet-cladding mechanical interaction (PCMI) and high cladding stresses that are strongly related to stressed corrosion cracking (SCC) which can lead to fuel rod failure. This work aims to quantify the stress fields in cladding with a focus on multiaxial stress relaxation response from an imposed strain and the resulting microstructure changes in textured zircaloy claddings. Stress relaxation tests resulting in multiaxial stress fields are performed to measure the material response of Zircaloy-4 nuclear cladding. Multiple experiments using samples from the same piece of cladding are performed so samples with the same processing history may be removed at strategic points throughout mechanical testing for microstructure characterization. The microstructure evolution is analyzed with mechanical response to evaluate the static recovery impact on crystallographic texture and grain morphology. Experiments are performed at Idaho National Laboratory using an experimental device capable of applying internal pressure and independent axial load to commercial nuclear cladding at elevated temperatures with stress- or strain-controlled experimental capability. X-ray diffraction and electron backscatter diffraction are used for microstructure analysis. Rheological models are used to visualize the multiaxial stress-strain relationships and isolate elastic, plastic, and viscoplastic material properties for analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

He Leak Testing of 3013 Inner Container Lids

The 3013-container package is used for the long-term storage of plutonium bearing materials at Department of Energy (DOE) sites. The package consists of three nested stainless-steel containers: outer, inner and convenience, which are made of stainless steel. The requirements of the DOE 3013 Standarda, also entail the development of a surveillance program, tasked to ensure structural integrity of the outer container for at least 50 years. The Materials Identification and Surveillance Working Group (MIS-WG) is specified to be responsible for selecting the containers for the surveillance as well as analyzing results. The MIS-WG determined that corrosion is a potential mechanism for container degradation with the most credible type of failure to be stress corrosion cracking (SCC) through wall breach of the inner container closure weld region (ICCWR). A breach of the inner container has the potential of allowing corrosive gases to be released into the safety class outer container.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

SNF Interim Storage Canister Corrosion and Surface Environment Investigations (FY21 Status Report)

This progress report describes work performed during FY21 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of canister materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In FY21, modeling and experimental work was performed that further defined our understanding of the potential chemical and physical environment present on canister surfaces at both marine and inland sites. Research also evaluated the relationship between the environment and the rate, extent, and morphology of corrosion, as well as the corrosion processes that occur. Finally, crack growth rate testing under relevant environmental conditions was initiated.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

SEM/EBSD Characterization of Baffle-Former Bolts Designated for High Energy X-ray Microscopy Studies at the Advanced Photon Source

This report documents the microstructural and surface characterization of two high-fluence baffle-former bolts (BFBs) harvested from a commercial pressurized water reactor (PWR) after nearly four decades of service. These bolts experienced extreme conditions including high neutron fluence, elevated temperatures, corrosive primary water, and sustained mechanical loads, all of which contribute to degradation mechanisms such as irradiation-assisted stress corrosion cracking (IASCC). Using scanning electron microscopy (SEM) and electron backscatter diffraction (EBSD), the study assessed bulk and near-surface microstructures, grain boundary characteristics, and in-service oxide layers. The results revealed significant differences between bolts exposed to 41 dpa and 29 dpa, including differences in recrystallization extent, deformation structures, and oxidation morphology. These findings establish critical baseline data for integration with advanced synchrotron-based high-energy X-ray microscopy (HEXM) and computational crystal plasticity models. Ultimately, the work supports improved mechanistic understanding and predictive capability regarding the long-term performance of reactor core internal components.

Lach, Tim [Oak Ridge National Laboratory (ORNL), O↗
NASA NTRS1980

The contribution of activated processes to Q

The possible role of activated processes in seismic attenuation is investigated. In this study, a solid is modeled by a parallel and series configuration of dashpots and springs. The contribution of stress and temperature activated processes to the long term dissipative behavior of this system is analyzed. Data from brittle rock deformation experiments suggest that one such process, stress corrosion cracking, may make a significant contribution to the attenuation factor, Q, especially for long period oscillations under significant tectonic stress.

Spetzler, H. A.↗
NASA NTRS1992

UPT scenarios: Implications for system reliability

The objective of this project was to examine the corrosion resistance of 316L stainless steel in several urine pre-treat solutions. Four solutions were examined: untreated urine (control); urine pretreated with oxone (potassium peroxymonosulfate sulfate); urine pretreated with sodium hypochlorite (NaOCl); and urine pretreated with ozone (O3). In accordance with current procedures, all solutions but the control were acidified to a pH of 2.5 using sulfuric acid--this suppresses the generation of ammonia in the solutions and is intended to limit microbial growth. Welded and unwelded coupons were exposed to each solution. In addition, Titanium coupons (welded and unwelded) were exposed to biologically active environmental control and life support system (ECLSS) water. Microbial attachment and biofilm growth were monitored. Ozone was examined as a biocide/oxidizer/corrosion preventative (simultaneous addition) and as a remediation method (added one week after exposure). In an unrelated effort, HP 9-4-30 coupons were exposed to biologically active solutions. Corrosion rates for welded and unwelded samples were determined--results were correlated to the ongoing HP 9-4-30 weldment stress corrosion study.

Walsh, Daniel W.↗
DOE OSTI2022

Evaluation of Applied Stress on Atmospheric Corrosion and Pitting Characteristics in 304L Stainless Steel

For this study, the effects of applied stress, ranging from tensile to compressive, on the atmospheric pitting corrosion behavior of 304L stainless steel (SS304L) were analyzed through accelerated atmospheric laboratory exposures and microelectrochemical cell analysis. After exposing the lateral surface of a SS304L four-point bend specimen to artificial seawater at 50°C and 35% relative humidity for 50 d, pitting characteristics were determined using optical profilometry and scanning electron microscopy. The SS304L microstructure was analyzed using electron backscatter diffraction. Additionally, localized electrochemical measurements were performed on a similar, unexposed, SS304L four-point bend bar to determine the effects of applied stress on corrosion susceptibility. Under the applied loads and the environment tested, the observed pitting characteristics showed no correlation with the applied stress (from 250 MPa to -250 MPa). Pitting depth, surface area, roundness, and distribution were found to be independent of location on the sample or applied stress. The lack of correlation between pitting statistics and applied stress was more likely due to the aggressive exposure environment, with a sea salt loading of 4 g/m 2 chloride. The pitting characteristics observed were instead governed by the available cathode current and salt distribution, which are a function of sea salt loading, as well as pre-existing underlying microstructure. In microelectrochemical cell experiments performed in Cl - environments comparable to the atmospheric exposure and in environments containing orders of magnitude lower Cl - concentrations, effects of the applied stress on corrosion susceptibility were only apparent in open-circuit potential in low Cl - concentration solutions. Cl - concentration governed the current density and transpassive dissolution potential.

36 MATERIALS SCIENCE↗
DOE OSTI2023

FY23 Status Report: SNF Interim Storage Canister Corrosion and Surface Environment Investigations

Work evaluating spent nuclear fuel (SNF) dry storage canister surface environments and canister corrosion progressed significantly in FY23, with the goal of developing a scientific understanding of the processes controlling initiation and growth of stress corrosion cracking (SCC) cracks in stainless steel canisters in relevant storage environments. The results of the work performed at Sandia National Laboratories (SNL) will guide future work and will contribute to the development of better tools for predicting potential canister penetration by SCC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1982

Dynamic fatigue of a machinable glass-ceramic

To assess the stress corrosion susceptibility of a machinable glass-ceramic, its dynamic fatigue behavior was investigated by measuring its strength as a function of stress rate. Fracture mechanics techniques were used to analyse the results for the purpose of making lifetime predictions for components of this material. This material was concluded to have only moderate resistance to stress in ambient conditions. The effects of specimen size on strength were assessed for the material used in this study: it was concluded that the Weibull edge-flaw scaling law adequately describes the observed strength-size relationship.

Smyth, K. K.↗
DOE OSTI2016

Selection of 3013 Containers for Field Surveillance. Fiscal Year 2016 Update

This update is the eighth in a series of reports that document the binning and sample selection of 3013 containers for the Field Surveillance program as part of the Integrated Surveillance Program. This report documents changes made to both the container binning assignments and the sample selection approach. Binning changes documented in this update are a result of changes to the prompt gamma calibration curves and the reassignment of a small number of Hanford items from the Pressure bin to the Pressure and Corrosion (P&C) bin. Field Surveillance sample selection changes are primarily a result of focusing future destructive examinations (DEs) on the potential for stress corrosion cracking in higher moisture containers in the P&C bin. The decision to focus the Field Surveillance program on higher moisture items is based on findings from both the Shelf-life testing program and DEs.

42 ENGINEERING↗
DOE OSTI2015

Gap Analysis to Support Modeling the Long-Term Degradation of Used Nuclear Fuel Canisters

Welded stainless steel canisters are being used worldwide for dry storage of used nuclear fuel (UNF) assemblies, and the number of canisters in use is steadily increasing. In support of work currently being pursued at Pacific Northwest National Laboratory to understand the atmospheric corrosion behavior of spent fuel dry storage systems, a gap analysis is underway to assess the state of knowledge for modeling of the long-term degradation of a UNF canister. The fundamental aim of this work is to inform research and development (R&D) efforts to establish a sound technical basis to support the extended dry storage of UNF for 100+ years. The analysis is considering all major components of the atmosphere corrosion degradation processes, ranging from contaminant sources and climatic interactions to regional conditions of particle transport and deposition, to microscale effects leading to stress corrosion cracking. The results of this gap analysis will be used to define the R&D pathway to develop an integrated multi-scale atmospheric corrosion modeling capability for UNF in dry storage canisters that can support the safe and reliable performance of these structures for more than 100 years.

Nuclear Fuel Canisters↗
DOE OSTI2024

Advanced Characterization Capabilities for Nuclear Materials via Nuclear Science User Facilities (NSUF)

Advanced post-irradiation examination (PIE) techniques are required to design new or improved nuclear materials, characterize, and understand in-core behavior of fuel and materials, and support the qualification of new reactor materials. The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy's only designated nuclear science user facility. NSUF provides researchers access to PIE capabilities at Idaho National Laboratory and at a diverse mix of university, national laboratory and industrial partner institutions. The PIE capabilities include novel destructive and non-destructive techniques for radiation damage characterization, such as advanced diffraction techniques (X-ray, electron, or neutron) coupled to extreme environments; in-situ observation of microstructural evolution under irradiation; in-situ irradiation to monitor corrosive attack in coolant environments; in-situ irradiation and mechanical testing; and test methods for synergistic effects of superimposed extreme environments (temperature, irradiation, stress, corrosion) on materials behaviors. This talk will provide an overview of NSUF PIE capabilities.

36 MATERIALS SCIENCE↗