Materials Research - Reports
| Report Title | Brief Narrative | Link | Date |
|---|---|---|---|
| Survey of Aging and Monitoring Concerns for Cables and Splices Due to Cable Repair and Replacement, PNNL-34576 | As portions of existing electrical cable runs in NPPs are replaced over time due to localized events, the total number of splices in NPPs are expected to increase. Relative to electrical cables, the body of knowledge regarding aging of splices and of splices in combination with aging cables in nuclear service environments in long-term operation is low. | Download | 7/31/2023 |
| Frequency Domain Reflectometry (FDR) Simulation Techniques for Digital Twin Representation of an Electrical Cable, Milestone M3LW-23OR0404023 | Simulation of nuclear electric cable system response to FDR tests can be instrumental to understanding the results of these tests and the nature and influence of various cable anomalies on test signatures. | Download | 4/14/2023 |
| Dose Rate Effects on Degradation of Nuclear Power Plant Electrical Cable Insulation at a Common Dose. PNNL-34068 | Dose rate effects (DRE) are an identified knowledge gap in relating accelerated aging of nuclear electrical cables to service aging, and here refer to gamma radiation-induced polymer degradation being a function of dose rate in addition to total absorbed dose. In this work, cross-linked polyethylene (XLPE) and ethylene propylene diene elastomer (EPDM) materials were subjected to accelerated aging at ambient temperature (26°C) at different gamma dose rates of 100, 200 and 1800 Gy/h for select exposure durations to achieve constant total doses of 170, 210 and 300 kGy to evaluate DRE. | Download | 3/24/2023 |
| Inverse Temperature Effects in Nuclear Power Plant Electrical Cable Insulation, PNNL-33296 | The results described in this report do not support the conclusion that ITE in cable qualification necessarily excludes safe continued use of existing cables. Inverse temperature effects were found to differ based on insulation material and on property measured. The current industry practice of subjecting cables to thermal aging followed by radiation aging at room temperature in qualification appears to be a conservative scenario for materials exhibiting ITE. Ongoing non-destructive cable system condition monitoring is encouraged to support repair and replace decisions for continued safe and effective use of electrical cables in long term operation. | Download | 9/9/2022 |
| Nondestructive Evaluation (NDE) of Cable Moisture Exposure using Frequency Domain Reflectometry (FDR), PNNL-31934 | Nuclear power facilities have experienced various electrical cable failures related to water exposure. The current industry response involves actions to de-water cable vaults, manholes, and other cable locations. These efforts require considerable expenditure of resources, which makes it desirable for the industry to have information on cable condition and history regarding their submergence and water exposure. | Download | 9/16/2021 |
| Inhomogeneous Aging of Nuclear Power Plant Electrical Cable Insulation, PNNL-31443 | Nuclear cable insulation samples of the two most common types, EPR and XPLE, from three of the most sourced manufacturers, Anaconda, Rockbestos, and Brand-Rex, were subjected to thermal aging at temperatures like those used in historic environmental qualification, 121, 136, 150, and 165°C. The aged materials were characterized for effects of aging using the conventional tensile EAB for all three and the indenter modulus for the EPR.Nuclear cable insulation samples of the two most common types, EPR and XPLE, from three of the most sourced manufacturers, Anaconda, Rockbestos, and Brand-Rex, were subjected to thermal aging at temperatures like those used in historic environmental qualification, 121, 136, 150, and 165°C. The aged materials were characterized for effects of aging using the conventional tensile EAB for all three and the indenter modulus for the EPR. | Download | 6/25/2021 |
| Sequential Versus Simultaneous Aging of XLPE and EPDM Nuclear Cable Insulation Subjected to Elevated Temperature and Gamma Radiation (Final Results), PNNL-30041 | This report addresses one of the knowledge gaps identified for the prediction of nuclear electrical cable aging: the conservatism of accelerated aging performed during historical cable qualification. If synergistic effects—degradation mechanisms unique to simultaneous exposure to thermal and gamma radiation stress—occur during cable service in an operating power plant, but not in the sequential thermal and radiation stress commonly used to simulate aging in the laboratory, then the aging case used in qualification may not be sufficiently conservative to envelope service aging. | Download | 12/12/2020 |
| Potential Life Extension Strategies for In-Service Degraded Cables, PNNL-30402 | Electrical cable systems are integral to the safe and efficient operation of nuclear power plants. As cable materials age over time in service, the performance and reliability of cables decrease and can fall to unacceptable levels. Effective techniques for assessing cable health and monitoring their condition are essential for determining cable status, predicting remaining useful life, and informing operators of the need for additional monitoring or replacement. | Download | 9/21/2020 |
| Cable Nondestructive Examination Online Monitoring for Nuclear Power Plants, PNNL-155612 | This PNNL report describes an investigation into cable inspection and monitoring methods that may be adapted to nuclear power plant online monitoring. There are numerous NDE methods to assess the condition of power plant cables. Most of these methods, however, require the cable systems to be offline and, in many cases, separated from their load. | Download | 9/20/2020 |
| Sequential Versus Simultaneous Aging of XLPE and EPDM Nuclear Cable Insulation Subjected to Elevated Temperature and Gamma Radiation, PNNL-30041 | This report addresses one of the knowledge gaps identified for the prediction of nuclear electrical cable aging, accelerated aging, performed during cable qualification. Synergistic effects are defined as polymer aging mechanisms specific to simultaneous or concurrent application of thermal and gamma radiation. If these effects differ from those experienced during sequentially applied thermal and gamma radiation stressors, then qualification that relies on sequential aging, as most historically did, may not well represent actual service aging when stressors exist simultaneously. | Download | 6/16/2020 |
| Dielectric Spectroscopy for Bulk Condition Assessment of Cable Insulation, PNNL-29092 | This report describes progress to date on the investigation of nondestructive test methods focusing on bulk cable insulation testing using a dielectric spectroscopy (DS) approach. This report addresses relevant literature coupled with a discussion of the theory of DS measurements and work on modeling to appreciate the influence of damage/defect profile on the bulk total cable measurement. | Download | 9/14/2019 |
| Evaluation of Inverse Temperature Effects on Cable Insulation Degradation in Accelerated Aging of High Priority Cable Insulation Materials, PNNL-29051 | Electrical cable insulation may degrade over time in service as a result of exposure to elevated temperature and gamma irradiation if the cable is at a location in the plant, such as inside containment, where both stresses occur. Cable polymer insulation generally degrades faster at higher temperatures. The extent of degradation is also generally proportional to the absorbed radiation dose. Accelerated aging at high temperatures and high doses has been used in the laboratory to simulate cable material aging that occurs in the milder temperatures and lower doses experienced by cables in service over the 40-year or longer life of the reactor. | Download | 8/18/2019 |
| Assessment of electrical breakdown strength relative to mechanical properities in insulations from harvested cable systems from nuclear power plants, ORNL SPR-2019/1145 | This report summarizes the characterization of the electrical breakdown strength in harvested electrical I&C cable insulation as a function of aging and its relationship to other mechanical properties. | Download | 4/12/2019 |
| Report on Initial Evaluations of Effects of Diffusion Limited Oxidation (DLO) Testing, PNNL-28351 | The objective of this study is to experimentally determine the thresholds of significant degradation in the most common nuclear cable insulation materials used by major cable manufacturers. Specifically, the focus is on thermal aging (temperature) and gamma irradiation aging (dose rate) thresholds. | Download | 12/18/2018 |
| Investigation of Thermal Aging Behavior for Harvested Crosslinked, PNNL-27729 | Investigation of Thermal Aging Behavior for Harvested Crosslinked Polyethylene and Ethylene-Propylene Rubber Cable Insulation, PNNL-27729, L. Fifield, M. Correa, Y. Shin, A. Zwoster, July 2018. | Download | 7/29/2018 |
| Interdigital Capacitance Local Non-Destructive Examination of Nuclear Power Plant Cable for Aging Management Programs, PNNL-27546 | Interdigital Capacitance Local Non-Destructive Examination of Nuclear Power Plant Cable for Aging Management Programs, PNNL-27546, S. Glass, L. Fifield, N. Bowler, A. Sriraman, W. Palmer, May 2018. | Download | 5/18/2018 |
| Summary Report on Activation Energies of Harvested Boston Insulated Wire and Okonite Cable Materials, ORNL SPR-2019/1145 | Summary Report on Activation Energies of Harvested Boston Insulated Wire and Okonite Cable Materials, M3LW-18OR0404015, R. Duckworth, May 2018. | Download | 5/18/2018 |
| Implementation of Concrete Creep Model in Grizzly, ORNL/TM-2017/729 | Implementation of Concrete Creep Model in Grizzly, ORNL/TM-2017/729, A. Giorla, November 2017. | Download | 11/18/2017 |
| Crystal River 3 Cable Materials for Thermal and Gamma Radiation Aging, PNNL-26785 | Crystal River 3 Cable Materials for Thermal and Gamma Radiation Aging, PNNL-26785, L. Fifield, M. Correa, A. Zwoster, September 2017. | Download | 9/20/2017 |
| Interdigital Capacitance Local Non-Destructive Examination of Nuclear Power Plant Cable for Aging Management Programs – Interim Report, PNNL-26087 | Interdigital Capacitance Local Non-Destructive Examination of Nuclear Power Plant Cable for Aging Management Programs – Interim Report, PNNL-26087, S. Glass, A. Jones, L. Fifield, M. Larche, N. Bowler, A. Sriraman, W. Palmer, September 2017. | Download | 9/20/2017 |
| Accelerated Thermal Aging Of Harvested Zion Electrical Cable Jacket And Insulation (Interim Report), M3LW-17OR0404110 | Accelerated Thermal Aging Of Harvested Zion Electrical Cable Jacket And Insulation (Interim Report), M3LW-17OR0404110, R. Duckworth, July 2017. | Download | 7/1/2017 |
| Analysis of simultaneous thermal/gamma radiation aging of cross-linked polyethylene (XLPE) insulation—interim status report, PNNL-26554 | Analysis of simultaneous thermal/gamma radiation aging of cross-linked polyethylene (XLPE) insulation—interim status report, PNNL-26554, L. Fifield, M. Correa, June 2017. | Download | 6/1/2017 |
| Physics-Based Modeling of Cable Insulation Conditions for Frequency Domain Reflectometry (FDR), PNNL-26493 | Physics-Based Modeling of Cable Insulation Conditions for Frequency Domain Reflectometry (FDR), PNNL-26493, S. Glass, A. Jones, L. Fifield, T. Hartman, N. Bowler, May 2017. | Download | 5/18/2017 |
| Update on Combined Thermal/Radiation Aging at Five Dose Rates in Chlorosulfonated Polyethylene (Hypalon)/Ethylene-Propylene Rubber (EPR) Cable Jacket Insulation System, M3LW-17OR0404019 | Update on Combined Thermal/Radiation Aging at Five Dose Rates in Chlorosulfonated Polyethylene (Hypalon)/Ethylene-Propylene Rubber (EPR) Cable Jacket Insulation System, M3LW-17OR0404019, R. Duckworth, March 2017. | Download | 3/18/2017 |
| Bulk and Distributed Electrical Cable Non-Destructive Examination Methods for Nuclear Power Plant Cable Aging Management Programs, PNNL-25634 | Bulk and Distributed Electrical Cable Non-Destructive Examination Methods for Nuclear Power Plant Cable Aging Management Programs, PNNL-25634, S. Glass, A. Jones, L. Fifield, T. Hartman, September 2016. | Download | 9/1/2016 |
| Status Report and Research Plan for Cables Harvested from Crystal River Unit 3 Nuclear Generating Plant, PNNL-25833 | Status Report and Research Plan for Cables Harvested from Crystal River Unit 3 Nuclear Generating Plant, PNNL-25833, L. Fifield, September 2016. | Download | 9/1/2016 |
| Progress in Characterizing Thermal Degradation of Ethylene-Propylene Rubber, PNNL-25713. | Progress in Characterizing Thermal Degradation of Ethylene-Propylene Rubber, PNNL-25713, L. Fifield, Q. Huang, M. Childers, M. Correa, Y. Shin, A. Zwoster, August 2016 | Download | 8/1/2016 |
| Progress in Characterizing Naturally-Aged Nuclear Power Plant Cables, PNNL-25439 | Progress in Characterizing Naturally-Aged Nuclear Power Plant Cables, PNNL-25439, L. Fifield, Q. Huang, M. Ian Childers, A. Zwoster, May 2016. | Download | 5/1/2016 |
| Evaluation of Localized Cable Test Methods for Nuclear Power Plant Cable Aging Management Programs, PNNL-25432 | Evaluation of Localized Cable Test Methods for Nuclear Power Plant Cable Aging Management Programs, PNNL-25432, S. Glass, L. Fifield, T. Hartman, May 2016. | Download | 5/1/2016 |
| Characterizing oxidation of cross-linked polyethylene and ethylene propylene rubber insulation materials by differential scanning calorimeter, PNNL-25172 | Characterizing oxidation of cross-linked polyethylene and ethylene propylene rubber insulation materials by differential scanning calorimeter, L. Fifield, J. Liu, Q. Huang, A. Zwoster, PNNL-25172, January 2016. | Download | 1/3/2016 |
| State-of-the-Art Assessment of NDE Techniques for Aging Cable Management in Nuclear Power Plants FY2015, PNNL-24649 | State-of-the-Art Assessment of NDE Techniques for Aging Cable Management in Nuclear Power Plants FY2015, S. Glass, L. Fifield, G. Dib, J. Tedeschi, A. Jones, T. Hartman, PNNL-24649, September 2015. | Download | 9/8/2015 |
| Progress on Analysis of Inverse Temperature Effects, Submerged Cables, Diffusion Limited Oxidation and Dose Rate Effects PNNL-24634 | Progress on Analysis of Inverse Temperature Effects, Submerged Cables, Diffusion Limited Oxidation and Dose Rate Effects. L.S. Fifield, M. P. Westman, M. K. Murphy, A. J. Zwoster, PNNL-24634, September 2015. | Download | 9/1/2015 |
| Assessment of Additional Key Indicators of Aging Cables in Nuclear Power Plants – Interim Status for FY 2015, PNNL-24649 | Assessment of Additional Key Indicators of Aging Cables in Nuclear Power Plants – Interim Status for FY 2015, P. Ramuhalli, L. Fifield, M. Prowant, G. Dib, J. Tedeschi, J. Suter, A. Jones, M. Good, S. Glass, A. Pardini, PNNL-24309, Pacific Northwest National Laboratory, May 2015. | Download | 5/1/2015 |
| Scoping/Design Study on Optimum Configuration for Combined Thermal/Radiation Aging of Cable Insulation Samples at ORNL | Scoping/Design Study on Optimum Configuration for Combined Thermal/Radiation Aging of Cable Insulation Samples at ORNL, R. Duckworth, Oak Ridge National Laboratory, March 2015. | Download | 3/1/2015 |
| Thermal aging modeling and validation on the Mo containing Fe-Cr-Ni alloys, ORNL/TM-2015/93 | Thermal aging modeling and validation on the Mo containing Fe-Cr-Ni alloys, Y. Yang, L. Tan, and J. Busby, ORNL/TM-2015/93, Oak Ridge National Laboratory, March 2015. | Download | 3/1/2015 |
| New Technologies for Repairing Aging Cables in Nuclear Power Plants | New Technologies for Repairing Aging Cables in Nuclear Power Plants, K. Simmons, L. Fifield, M. Westman, and J. Roberts, Pacific Northwest National Laboratory, September 2014. | Download | 9/1/2014 |
| Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants—Interim Status for FY 2014, PNNL-23624 | Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants—Interim Status for FY 2014, K. Simmons, L. Fifield, M Westman, J. Tedeschi, A. Jones, M. Prowant, A. Pardini, and P. Ramuhalli, Pacific Northwest National Laboratory, PNNL-23624, September 2014. | Download | 9/1/2014 |
| Preliminary List of Aging Conditions and Measurement Methods to be Examined for Key Indicators of Cable Aging – Status Summary | Preliminary List of Aging Conditions and Measurement Methods to be Examined for Key Indicators of Cable Aging – Status Summary, K. Simmons, PNNL, April 2013. | Download | 5/1/2013 |
| Nondestructive Examination (NDE) Detection and Characterization of Degradation Precursors, PNNL-21692, P. Ramuhalli J.W. Griffin, R.M. Meyer, S.G. Pitman, J.M. Fricke, M.E. Dahl, M.S. Prowant, T.A. Kafentzis, J.B. Coble, T.J. Roosendaal, September 2012. | Nondestructive Examination (NDE) Detection and Characterization of Degradation Precursors, PNNL-21692, P. Ramuhalli J.W. Griffin, R.M. Meyer, S.G. Pitman, J.M. Fricke, M.E. Dahl, M.S. Prowant, T.A. Kafentzis, J.B. Coble, T.J. Roosendaal, September 2012. | Download | 9/1/2012 |
| Assessment of Opportunities for Acquiring Plant Materials to Aid in Model Validation, G. Von White and R. Bernstein, Sandia National Laboratory, June 2012. | Assessment of Opportunities for Acquiring Plant Materials to Aid in Model Validation, G. Von White and R. Bernstein, Sandia National Laboratory, June 2012. | Download | 6/1/2012 |
| Status of Cable Aging Knowledge Gaps Identified in the Expanded Materials Degradation Assessment (EMDA) | Knowledge gaps identified in the Expanded Materials Degradation Assessment (EMDA), Volume 5: Aging of Cables and Cable Systems represent concerns that the assumptions made in 40-year environmental qualification of cables may be weak, that the pre-aging of cables prior to LOCA testing may have represented less than 40-year equivalence, and that consequently the EQ process may not be conservative and thereby overpredict cable useful lifetime. | Download | 09/15/2023 |
| Methodological guideline for industry focusing on characterization procedures to assess the risk of irradiation degradation of concrete in the biological shield, ORNL/TM-2024/3287 | This report presents a methodological guideline for industry to assess potential irradiation damage of their concrete formulation at extended operation. It describes experimental methods that can be used to characterize the chemical and physical properties of unirradiated and irradiated aggregates. | Download | 04/24/2024 |
| Complete the evaluation of the effect of 60-year aging on the susceptibility to stress corrosion cracking of a first layer Alloy 152 butter weld deposited on low alloy steel, ANL/LWRS-24/1 | Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). | Download | 09/14/2024 |
| Complete evaluation of the effect of dissolved oxygen on stress corrosion cracking initiation of austenitic stainless steel in PWR primary water | : About 90% of the primary components in pressurized water reactors are made of stainless steels. Operational experience with these stainless steels has generally been very good given their extensive use. Nevertheless, stress corrosion cracking (SCC) cases of unirradiated austenitic stainless steels in pressurized water reactor (PWR) primary water have occurred in the field since the early 1980s and continued to this day. | Download | 09/14/2024 |
| Complete the Second Weld Campaign on Ni Alloy 182 Using Optimized Welding Parameters and Complete the Initial Weld Quality Evaluation, ORNL/SPR-2024/3575 | This report summarizes the latest welding campaign on irradiated Ni-base alloy 182 and preliminary weld inspections at the Radiochemical Engineering Development Center (REDC). A collaborative effort between the U.S. Department of Energy, Electric Power Research Institute, and Oak Ridge National Laboratory, the campaign focused on refining laser parameters for welding boron-doped alloy 182 (14, 15, and 23 wppm). | Download | 09/14/2024 |
| Evaluation of Clamshell Current Coupler for Online Frequency Domain and Spread Spectrum Time Domain Reflectometry to Detect Anomalies in Energized Cables, PNNL-36530 | Safety-critical nuclear power plant cables were initially qualified for 40 years. However, as plants extend their operating licenses to 60 and 80 years, justification for continued safe operation includes test and monitoring programs. This will become more important as the industry moves to condition based qualification programs. Cable test programs traditionally involve manual interventions to disconnect cables, perform one or several tests, then reconnect the systems, usually during refueling outages occurring only every 18 to 24 months. | Download | 09/13/2024 |
| Assessment of Machine Learning for Ultrasonic Nondestructive Evaluation of Alkali–Silica Reaction in Concrete, ORNL/TM-2024/3295 | This report evaluates the effectiveness of two ML models [support vector regression (SVR) and deep neural network (DNN)] in predicting concrete material damage induced by ASR based on long-term ultrasonic monitoring data. The focus was to study the factors that may impact the ML model performance when using ultrasonic NDE data for concrete damage evaluation. A framework for using ML for NDE of concrete material properties based on ultrasonic data was also summarized in this work. | Download | 03/11/2024 |
| Report Title | Brief Narrative | Link | Date |
|---|---|---|---|
| Machine Learning for Processing Ultrasonic Data from Long-Term Monitoring of Concrete with Alkali-Silica Reaction (ASR) | This report provides a comparison of the performance of four machine learning models to establish a relationship between concrete expansion due to alkali-silica reaction (ASR) and ultrasonic signals. Expansion and ultrasonic data collected from two concrete blocks exposed to accelerated ASR were used. Wave velocity and 12 wavelet features were extracted from the ultrasonic data and used to train three different machine learning models: linear regression, support vector regression, and shallow neural network. Data from one of the blocks was used to train the models, while data from the other block to test them. These three models exhibited poor performance on the test data with prediction R2 values smaller than 0.71 and RMSE values larger than 0.09%. To improve the performance of these models, a feature selection process was implemented leading to choosing 6 features with the highest correlation coefficients with the expansion. This selection process led to better R2 and RMSE values for all three models and comparable performance for all of them. | Download | 6/29/2023 |
| Ultrasonic Model Based Iterative Reconstruction of Experimental Concrete Specimens at EPR, ORNL/SPR-2022/2318 | Concrete is a critical component of NPPs for both safety and reliability issues. The original plant designers could not have anticipated the extended lifespan these facilities may reach. Continuous monitoring of the concrete for signs of degradation through NDE may be applied to many levels of NPP infrastructure. | Download | 1/1/2022 |
| Development of a Reconstruction Methodology Based on X-Ray Computed Tomography to Generate Realistic 3D Concrete Microstructures in MOSAIC, ORNL/TM-2021/2156 | Despite being passive components, concrete structures represent a major capital investment in nuclear power plants. However, some concrete components are critical for the safety and long-term operation of the reactors: the concrete containment building protects the reactors from external aggression. The concrete biological shield (CBS) contains the radiation exiting the reactor to protect equipment and personnel. Depending on the design, the CBS also has a function for supporting the reactor’s systems. This function needs to be ensured in-service and during accident scenarios. | Download | 8/26/2021 |
| Assessment of Grizzly Capabilities for Reactor Pressure Vessels and Reinforced Concrete Structures, INL/EXT-20-00617 | Over the last several years, capabilities to simulate the progression and effects of degradation in critical structures in LWR nuclear power plants have been under development in the Grizzly code. Age-related material degradation is important for a number of systems in LWRs, but the main focus for Grizzly development has been on RPVs and concrete structures because of their central role and the difficulty of replacement of these structures if they are found to be degraded to an unacceptable degree. | Download | 9/21/2020 |
| Comparative Analysis of Nondestructive Examination Techniques of Enhanced Model Based Iterative Reconstruction (MBIR) and Frequency-banded Synthetic Aperture Focusing Technique (SAFT) Reconstructions, ORNL/SPR-2019/1240 | As the existing US fleet of NPPs approach their expected lifetimes and only a few new reactors are coming online, the industry is looking to extend their licenses beyond 60 years and improve efficiencies through research-informed aging management programs. Reinforced concrete is an inexpensive, strong material widely used in the nuclear industry. | Download | 9/11/2019 |
| Two-Modulator Generalized Ellipsometry Microscope, ORNL/TM-2019/1025 | The examination of microstructures in geological materials, including aggregates used in concrete, is a complex task. Natural rocks contain crystals of varying sizes, ranging from submicron to centimeters, and these crystals can have different chemical compositions and orientations. | Download | 6/1/2019 |
| Microstructural characterization of the alkali-silica reaction (ASR)-induced damage in structural concrete test blocks at the University of Tennessee, Knoxville, ORNL/SPR-2019/1139 | The key findings and conclusions from phase (1) can be summarized as follows: | Download | 5/28/2019 |
| Identification of Mechanisms to Study Alkali-Silica Reaction Effects on Stress-Confined Concrete Nuclear Thick Structures: Interpretation of the Complete Monitoring Data and Nondestructive Evaluation of the Alkali-Silica Reaction Test Assembly, ORNL/SPR-2 | The evaluation of the structural impact of alkali-silica reaction (ASR) on concrete in nuclear power plants is not… | Download | 8/27/2018 |
| Development of Fast Fourier Transform (FFT) micro-mechanical simulations of concrete specimens characterized by micro-X-ray fluorescence Alain, ORNL/TM-2017/367 | Development of Fast Fourier Transform (FFT) micro-mechanical simulations of concrete specimens characterized by micro-X-ray fluorescence Alain, ORNL/TM-2017/367, A. Giorla, August 2017. | Download | 8/18/2017 |
| IMAC Database v.0.3. – Concrete, ORNL/SPR-2017/436 | IMAC Database v.0.3. – Concrete, ORNL/SPR-2017/436, Y. Pape, August 2017. | Download | 8/18/2017 |
| Linear Array Ultrasonic Testing Of A Thick Concrete Specimen For Nondestructive Evaluation, ORNL/TM-2017/156 | Linear Array Ultrasonic Testing Of A Thick Concrete Specimen For Nondestructive Evaluation, ORNL/TM-2017/156, D. Clayton, N. Ezell, L. Khazanovich, M. Zammerachi, April 2017. | Download | 5/29/2017 |
| IMAC Database v.0.1. Minerals, ORNL/TM-2016/ORNL/TM-2016/753 | IMAC Database v.0.1. – Minerals, ORNL/TM-2016/ORNL/TM-2016/753, Y. Pape, December 2016. | Download | 12/2/2016 |
| Independent Modeling of the Alkali-Silica Reaction: Mock-up Test Block, ORNL/TM-2016/537 | Independent Modeling of the Alkali-Silica Reaction: Mock-up Test Block, ORNL/TM-2016/537, M. Hariri-Ardebili, V. Saouma, Y. LePape, September 15, 2016. | Download | 9/15/2016 |
| Simulation of Concrete Members Affected by Alkali-Silica Reaction with Grizzly, ORNL/TM-2016/523 | Simulation of Concrete Members Affected by Alkali-Silica Reaction with Grizzly, ORNL/TM-2016/523, A. Giorla, September 2016. | Download | 9/1/2016 |
| Evaluation of Advanced Signal Processing Techniques to Improve Detection and Identification of Embedded Defects, ORNL-TM-2016-482 | Evaluation of Advanced Signal Processing Techniques to Improve Detection and Identification of Embedded Defects, ORNL-TM-2016-482, D. Clayton, H. Santos-Villalobos, J. Baba, September 2016. | Download | 9/1/2016 |
| Identification of Mechanisms to Study Alkali-Silica Reaction Effects on Stressed-Confined Concrete Nuclear Thick Structures | Identification of Mechanisms to Study Alkali-Silica Reaction Effects on Stressed-Confined Concrete Nuclear Thick Structures, Z. Ma, S. Pape, N. Hayes, Q. Gui, A. Elhassan, Y. Jing, August 31, 2016. | Download | 8/31/2016 |
| LWRS contribution to the RILEM benchmark on materials modeling of ASR – Preliminary results, ORNL/LTR-2016/395 | LWRS contribution to the RILEM benchmark on materials modeling of ASR – Preliminary results, ORNL/LTR-2016/395, Y. Pape, A. Gloria, August 2016. | Download | 8/1/2016 |
| Linear Array Ultrasonic Test Results from Alkali-Silica Reaction (ASR) Specimens, ORNL-TM-2016-159 | Linear Array Ultrasonic Test Results from Alkali-Silica Reaction (ASR) Specimens, ORNL-TM-2016-159, D. Clayton, L. Khazanovich, L. Salles, April 2016. | Download | 4/1/2016 |
| Alkali-Silica Reaction test Assembly – Report describing the procurement of materials and equipment for the ASR test assembly Identification of Mechanisms to Study Alkali-Silica Reaction Effects on Stressed-confined Concrete Nuclear Thick Structures | Alkali-Silica Reaction test Assembly – Report describing the procurement of materials and equipment for the ASR test assembly Identification of Mechanisms to Study Alkali-Silica Reaction Effects on Stressed-confined Concrete Nuclear Thick Structures, Z Ma, University of Tennessee, January 2016. | Download | 1/15/2016 |
| LWRS Nondestructive Evaluation (NDE) for Concrete Research and Development Roadmap, ORNL/TM-2012/360, D. Clayton, M. Hileman, September 2012. | LWRS Nondestructive Evaluation (NDE) for Concrete Research and Development Roadmap, ORNL/TM-2012/360, D. Clayton, M. Hileman, September 2012. | 9/1/2012 | |
| LWRS Concrete NDE Workshop Summary, D.A. Clayton, August 9, 2012. | LWRS Concrete NDE Workshop Summary, D.A. Clayton, August 9, 2012. | Download | 8/9/2012 |
| Assessment of Machine Learning for Ultrasonic Nondestructive Evaluation of Alkali–Silica Reaction in Concrete, ORNL/TM-2024/3295 | This report evaluates the effectiveness of two ML models [support vector regression (SVR) and deep neural network (DNN)] in predicting concrete material damage induced by ASR based on long-term ultrasonic monitoring data. The focus was to study the factors that may impact the ML model performance when using ultrasonic NDE data for concrete damage evaluation. A framework for using ML for NDE of concrete material properties based on ultrasonic data was also summarized in this work. | Download | 3/11/2024 |
| Ultrasonic Model Based Iterative Reconstruction of Experimental Concrete Specimens at EPR, ORNL/SPR-2022/2318 | Concrete is a critical component of NPPs for both safety and reliability issues. The original plant designers could not have anticipated the extended lifespan these facilities may reach. Continuous monitoring of the concrete for signs of degradation through NDE may be applied to many levels of NPP infrastructure. | Download | 1/1/2022 |
| Reconstruction of 3D Concrete Microstructures Combining High-Resolution Characterization and Convolutional Neural Network for Image Segmentation | After water, concrete is the second most used material in the world. Concrete’s forming adaptability and low-cost constituents make it a predominant material used in the construction of civil infrastructures in nuclear power plants such as concrete biological shields, containment buildings, turbine buildings, fuel handling and storage buildings, underground piping for cooling, cooling towers. | Download | 6/12/2022 |
| Two-Modulator Generalized Ellipsometry Microscope, ORNL/TM-2019/1025 | The examination of microstructures in geological materials, including aggregates used in concrete, is a complex task. Natural rocks contain crystals of varying sizes, ranging from submicron to centimeters, and these crystals can have different chemical compositions and orientations. | Download | 6/1/2019 |
| Comparative Analysis of Nondestructive Examination Techniques of Enhanced Model Based Iterative Reconstruction (MBIR) and Frequency-banded Synthetic Aperture Focusing Technique (SAFT) Reconstructions, ORNL/SPR-2019/1240 | As the existing US fleet of NPPs approach their expected lifetimes and only a few new reactors are coming online, the industry is looking to extend their licenses beyond 60 years and improve efficiencies through research-informed aging management programs. Reinforced concrete is an inexpensive, strong material widely used in the nuclear industry. | Download | 9/11/2019 |
| Assessment of Grizzly Capabilities for Reactor Pressure Vessels and Reinforced Concrete Structures, INL/EXT-20-00617 | Over the last several years, capabilities to simulate the progression and effects of degradation in critical structures in LWR nuclear power plants have been under development in the Grizzly code. Age-related material degradation is important for a number of systems in LWRs, but the main focus for Grizzly development has been on RPVs and concrete structures because of their central role and the difficulty of replacement of these structures if they are found to be degraded to an unacceptable degree. | Download | 9/21/2020 |
| Development of a Reconstruction Methodology Based on X-Ray Computed Tomography to Generate Realistic 3D Concrete Microstructures in MOSAIC, ORNL/TM-2021/2156 | Despite being passive components, concrete structures represent a major capital investment in nuclear power plants. However, some concrete components are critical for the safety and long-term operation of the reactors: the concrete containment building protects the reactors from external aggression. The concrete biological shield (CBS) contains the radiation exiting the reactor to protect equipment and personnel. Depending on the design, the CBS also has a function for supporting the reactor’s systems. | Download | 8/26/2021 |
| Machine Learning for Processing Ultrasonic Data from Long-Term Monitoring of Concrete with Alkali-Silica Reaction (ASR) | This report provides a comparison of the performance of four machine learning models to establish a relationship between concrete expansion due to alkali-silica reaction (ASR) and ultrasonic signals. Expansion and ultrasonic data collected from two concrete blocks exposed to accelerated ASR were used. Wave velocity and 12 wavelet features were extracted from the ultrasonic data and used to train three different machine learning models: linear regression, support vector regression, and shallow neural network. Data from one of the blocks was used to train the models, while data from the other block to test them. These three models exhibited poor performance on the test data with prediction R2 values smaller than 0.71 and RMSE values larger than 0.09%. | Download | 6/29/2023 |
| Performance Comparison of Machine Learning Models for Ultrasonic Nondestructive Evaluation of Alkali-Silica Reaction in Concrete | This report presents a comparative analysis of four ML regression models for predicting concrete material damage induced by ASR expansion using long-term ultrasonic data monitoring. The models investigated include linear regression (LR), support vector regression (SVR), shallow neural networks (NN), and deep neural networks (DNN). LR, SVR, and shallow NN models use features extracted from ultrasonic signals, whereas the DNN model processes time-domain ultrasonic signals and frequency spectra directly. | Download | 08/01/2024 |
| Methodological guideline for industry focusing on characterization procedures to assess the risk of irradiation degradation of concrete in the biological shield, ORNL/TM-2024/3287 | his report presents a methodological guideline for industry to assess potential irradiation damage of their concrete formulation at extended operation. It describes experimental methods that can be used to characterize the chemical and physical properties of unirradiated and irradiated aggregates. | Download | 04/02/2024 |
| High neutron and gamma dose on effects on calcium silicate deuterate, ORNL/SPR-2024/3454 | With the planned prolonged operation of LWRs in the US, it is imperative to understand the effects of radiation in all plant components including the concrete biological shield, so that reliable projections of properties and performance can be estimated for this civil structure. The effects of combined neutron and gamma radiation on cement hydrates were investigated in this report. The combination of gamma and neutron radiation on cement hydrates has not been studied in detail before, as much past experimental research has focused on the effects of gamma, and some theoretical studies have focused on neutron effects. Thus, this report presents a novel experimental study combining both effects. | Download | 07/17/2024 |
| Implementation Plan and Initial Development of Nuclear Concrete Materials Database for Light Water Reactor Sustainability Program, W. Ren and D. Naus, ORNL, B. Oland, XCEL Engineering, ORNL/TM- 2010/177, September 2010. | Implementation Plan and Initial Development of Nuclear Concrete Materials Database for Light Water Reactor Sustainability Program, W. Ren and D. Naus, ORNL, B. Oland, XCEL Engineering, ORNL/TM- 2010/177, September 2010. | Download | 9/23/2010 |
| Report Title | Brief Narrative | Link | Date |
|---|---|---|---|
| Complete the first phase of the comprehensive characterization of repair welding performed on irradiated Ni alloy182 using stress improved laser welding in collaboration with EPRI,ORNL/TM-2023/3109 | This report describes the research activities conducted in FY 2023 on post-weld evaluation and characterization of the quality and properties of welds made on irradiated Ni-base alloy 182 (with boron concentration up to 15 wppm) by the advanced laser repair welding technology developed under the DOE Light Water Reactor Sustainability Program and EPRI LTO program. The research represents a major progress in repair welding of highly irradiated helium containing reactor internals and its feasibility to use the ABSI-LW technology developed in this program. | Download | 9/15/2023 |
| Comprehensive Characterization of Helium-induced Degradation of the Friction Stir Weld on Neutron Irradiated 304L Stainless Steel , ORNL/TM-2023/3088 | Fractography analysis showed mostly ductile fracture with some brittle areas, related, as believed, to the helium-rich spots. From the BM observation, such helium enriched spots might be produced artificially in the BM production but needs further characterization to determine. Helium-rich spots led to localized fracture effects along the specimen edges and probably in the bulk. Helium-related issues were the most pronounced in the top retreating side (TR) specimen, leading to delamination effects. | Download | 9/12/2023 |
| Development of Digital Twin Predictive Model for PWR Components: Updates on Multi Times Series Temperature Prediction Using Recurrent Neural Network, DMW Fatigue Tests, System Level Thermal-Mechanical-Stress Analysis. ANL/LWRS-21/02 | Friction stir welding of irradiated material in the 7930 hot cells at Oak Ridge National Laboratory was paused in 2020 after attempts to resolve the weld quality issues experienced in 2018-19. Plans have been developed but budget and manpower limitations have prevented further process development and evaluation of welding equipment as recommended in the 2020 milestone report. | Download | 9/17/2021 |
| Improve the design of and assess upgrades to the Friction Stir welding system to reduce defects on surrogate unirradiated materials, ORNL/TM-2021/2242 | Friction stir welding of irradiated material in the 7930 hot cells at Oak Ridge National Laboratory was paused in 2020 after attempts to resolve the weld quality issues experienced in 2018-19. Plans have been developed but budget and manpower limitations have prevented further process development and evaluation of welding equipment as recommended in the 2020 milestone report. | Download | 9/17/2021 |
| Microstructure and Mechanical Performance of the Friction Stir Welds Performed on Neutron-Irradiated Steel with Helium, ORNL/TM-2021/2079 | This report describes new experimental results on the microstructure and mechanical performance of the friction stir welds made on neutron-irradiated steel with helium. The report focuses on helium-related issues, specifically, helium-induced degradation in the welded joint, aiming to repair irradiated components of NPPs. | Download | 8/31/2021 |
| Conduct weld campaign (FY-21-1) on irradiated materials provided by the Canadian Nuclear Laboratory (CNL), including baseline post-weld evaluation and testing, ORNL/TM-2020/1673 | A collaborative research on developing advanced welding technologies for irradiated stainless steel 304 materials between CNL and ORNL has been established to support both U.S. and Canadian interests in evaluation of weld repair techniques on irradiated materials to support continuous operation of commercial nuclear power. The work utilizes unique material from the National Research Universal (NRU) reactor and the specialized welding hot cell facility at ORNL. | Download | 8/13/2021 |
| Intermediate-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at the LWR Relevant Temperature, ORNL/TM-2020/1754 | Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. | Download | 9/21/2020 |
| Steam Oxidation of Alloy 718a and Tensile Properties of Select Advanced Replacement Alloys for LWR Core Internals, ORNL/TM-2019/1308 | Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. | Download | 9/12/2019 |
| Recent Technological Advances In Welding Irradiated Austenitic Steel With Helium, ORNL/SPR-2019/1347 | This report describes the post-weld study of weld quality and weld properties of irradiated 304 stainless steel made using both Auxiliary Beam Stress Improved Laser Welding (ABSI-LW) and Friction Stir Welding (FSW) processes and conducted in FY 2019. The results demonstrate that both ABSI-LW and FSW processes, developed in this program, mitigated helium-induced cracking during weld repair of helium-containing irradiated SS304 stainless steels. | Download | 9/2/2019 |
| Develop Parameters and Characterize the Quality of Friction Stir and Laser Weld-Repaired, Irradiated Structural Materials Representative of Extended Reactor Service Life, ORNL/SPR-2019-1035 | This report summarizes the most recent welding campaign on irradiated 304L and 316L at the REDC, and post-welded quality and microstructure characterization, as well, as microhardness testing of ongoing weld campaigns using irradiated stainless-steel alloys at the LAMDA facilities at ORNL. | Download | 4/18/2019 |
| Metallurgical Aspects Influencing the Resistance to Steam Oxidation and Fracture Toughness of Select Advanced Replacement Alloys for LWR Core Internals, ORNL/TM-2018/973 | The corrosion resistance of materials in a water environment is a crucial requirement for core internal materials in light-water reactors (LWRs). High-temperature steam oxidation tests serve as an accelerated life testing method that not only reveals potential failure modes within a short duration but also enables the evaluation of materials’ resistance to accidental scenarios. These tests provide valuable insights into the performance of materials under extreme conditions and contribute to ensuring the reliability and safety of LWR core internal components. | Download | 9/7/2018 |
| Complete Report on Development of Weld Repair Technology M2LW-18OR0406014 | The focus of the research was on post-welding procedures for irradiated alloys, with an emphasis on high helium content stainless steel. The initial measurements of helium content and microstructure characterization indicated successful advanced laser and friction stir welding techniques. The ongoing efforts to weld irradiated alloys with high helium concentrations and thoroughly analyze the results aim to establish validated repair techniques and guidelines for the nuclear industry. | Download | 9/5/2018 |
| Report on the Progress of Weld Development of Irradiated Materials at the Oak Ridge National Laboratory, ORNL/SPR-2018/833 | Report on the Progress of Weld Development of Irradiated Materials at the Oak Ridge National Laboratory, Z. Feng, R. Miller, J. Chen, W. Tang, S. Clark, B. Gibson, M. Vance, G. Frederick, J. Tatman, B. Sutton, April 2018. | Download | 4/18/2018 |
| Proton Irradiation Screening Results of Select Advanced Replacement Alloys for Core Internals | Proton Irradiation Screening Results of Select Advanced Replacement Alloys for Core Internals, G. Was, M. Wang, M. Song, C. Lear, September 2017. | Download | 9/20/2017 |
| Fracture Toughness Evaluation of Select Advanced Replacement Alloys for LWR Core Internals, ORNL/TM-2017/377 | Fracture Toughness Evaluation of Select Advanced Replacement Alloys for LWR Core Internals, ORNL/TM-2017/377, X. Chen, L. Tan, August 25, 2017. | Download | 8/25/2017 |
| Report Summarizing the Status of Second Round Irradiation Experiments and Assessment of Materials Available for Testing Advanced Welding Techniques, M3LW-17OR0406013 | Report Summarizing the Status of Second Round Irradiation Experiments and Assessment of Materials Available for Testing Advanced Welding Techniques, Z. Feng, R. Miller, N. Cetiner, X. Hu, S. Clark, G. Frederick, B. Sutton, July 2017. | Download | 7/18/2017 |
| Report Summarizing the Effort Required to Initiate Welding of Irradiated Materials Within the Welding Cubicle, M3LW-17OR0406012 | Report Summarizing the Effort Required to Initiate Welding of Irradiated Materials Within the Welding Cubicle, Z, Feng, W. Tang, J. Chen, X. Hu, R. Miller, B. Gibson, A. Smith, M. Vance, S. Clark, G. Frederick, J. Tatman, B. Sutton, June 2017. | Download | 6/18/2017 |
| High-Temperature Steam Oxidation Testing of Select Advanced Replacement Alloys for Potential Core Internals, ORNL/TM-2017/228 | High-Temperature Steam Oxidation Testing of Select Advanced Replacement Alloys for Potential Core Internals, ORNL/TM-2017/228, L. Tan, B. Pint, May 19, 2017. | Download | 5/19/2017 |
| Assessment of Radiation Resistance of Selected Alloys in The ARRM Program Using Ion Irradiation, M3LW-16OR0406024 | Assessment of Radiation Resistance of Selected Alloys in The ARRM Program Using Ion Irradiation, G. Was, M. Song, M. Wang, September 30, 2016. | Download | 9/30/2016 |
| Toughness and High-Temperature Steam Oxidation Evaluations of Advanced Alloys for Core Internals, ORNL/TM-2016/371 | Toughness and High-Temperature Steam Oxidation Evaluations of Advanced Alloys for Core Internals, ORNL/TM-2016/371, September 16, 2016. | Download | 9/16/2016 |
| L3 Milestone M3LW-16OR0406015, Provide documentation on the information and states of both the first and second series of neutron irradiate series of B-doped steels for the weld validation testing to begin in FY17 | L3 Milestone M3LW-16OR0406015, Provide documentation on the information and states of both the first and second series of neutron irradiate series of B-doped steels for the weld validation testing to begin in FY17, Z. Feng, N. Cetiner, X. Hu, R. Miller, G. Frederick, B. Sutton, September 2016. | Download | 9/1/2016 |
| Report Detailing Friction Stir Welding Process Development for the Hot Cell Welding System | Report Detailing Friction Stir Welding Process Development for the Hot Cell Welding System, M3LW-16OR0406014, W. Tang, B. Gibson, Z. Feng, S. Clark, A. Peterson, G. Frederick, September 2016. | Download | 9/1/2016 |
| Report on the Installation of the Integrated Welding Hot Cell at ORNL Building 7930, M2LW-16OR0406013 | Report on the Installation of the Integrated Welding Hot Cell at ORNL Building 7930, M2LW-16OR0406013, Z. Feng, W. Tang, R. Miller, B. Gibson, A. Peterson, J. Tatman, G. Frederick, June 30, 2016. | Download | 6/30/2016 |
| Status update of advanced alloys for Advanced Radiation Resistant Materials Program | Status update of advanced alloys for Advanced Radiation Resistant Materials Program, L. Tan, D. Hoelzer, and J. Busby, Oak Ridge National Laboratory, March 2014. | Download | 3/1/2014 |
| Use of Computational Model to Design and Optimize Welding Conditions to Suppress Helium Cracking during Welding, W. Zhang Z. Feng and E. Willis, June 2012. | Use of Computational Model to Design and Optimize Welding Conditions to Suppress Helium Cracking during Welding, W. Zhang Z. Feng and E. Willis, June 2012. | Download | 6/1/2012 |
| Develop Baseline Computational Model for Proactive Welding Stress Management to Suppress Helium-Induced Cracking During Weld Repair, W. Zhang and Z. Feng, ORNL, September 2011. | Develop Baseline Computational Model for Proactive Welding Stress Management to Suppress Helium-Induced Cracking During Weld Repair, W. Zhang and Z. Feng, ORNL, September 2011. | Download | 9/1/2011 |
| Complete the first phase of the comprehensive characterization of repair welding performed on irradiated Ni alloy182 using stress improved laser welding in collaboration with EPRI,ORNL/TM-2023/3109 | This report describes the research activities conducted in FY 2023 on post-weld evaluation and characterization of the quality and properties of welds made on irradiated Ni-base alloy 182 (with boron concentration up to 15 wppm) by the advanced laser repair welding technology developed under the DOE Light Water Reactor Sustainability Program and EPRI LTO program. The research represents a major progress in repair welding of highly irradiated helium containing reactor internals and its feasibility to use the ABSI-LW technology developed in this program. | Download | 09/15/2023 |
| Report Title | Brief Narrative | Link | Date |
|---|---|---|---|
| Complete the first phase of the comprehensive characterization of repair welding performed on irradiated Ni alloy182 using stress improved laser welding in collaboration with EPRI,ORNL/TM-2023/3109 | This report describes the research activities conducted in FY 2023 on post-weld evaluation and characterization of the quality and properties of welds made on irradiated Ni-base alloy 182 (with boron concentration up to 15 wppm) by the advanced laser repair welding technology developed under the DOE Light Water Reactor Sustainability Program and EPRI LTO program. The research represents a major progress in repair welding of highly irradiated helium containing reactor internals and its feasibility to use the ABSI-LW technology developed in this program. | Download | 9/15/2023 |
| Comprehensive Characterization of Helium-induced Degradation of the Friction Stir Weld on Neutron Irradiated 304L Stainless Steel , ORNL/TM-2023/3088 | Fractography analysis showed mostly ductile fracture with some brittle areas, related, as believed, to the helium-rich spots. From the BM observation, such helium enriched spots might be produced artificially in the BM production but needs further characterization to determine. Helium-rich spots led to localized fracture effects along the specimen edges and probably in the bulk. Helium-related issues were the most pronounced in the top retreating side (TR) specimen, leading to delamination effects. | Download | 9/12/2023 |
| Development of Digital Twin Predictive Model for PWR Components: Updates on Multi Times Series Temperature Prediction Using Recurrent Neural Network, DMW Fatigue Tests, System Level Thermal-Mechanical-Stress Analysis. ANL/LWRS-21/02 | Friction stir welding of irradiated material in the 7930 hot cells at Oak Ridge National Laboratory was paused in 2020 after attempts to resolve the weld quality issues experienced in 2018-19. Plans have been developed but budget and manpower limitations have prevented further process development and evaluation of welding equipment as recommended in the 2020 milestone report. | Download | 9/17/2021 |
| Improve the design of and assess upgrades to the Friction Stir welding system to reduce defects on surrogate unirradiated materials, ORNL/TM-2021/2242 | Friction stir welding of irradiated material in the 7930 hot cells at Oak Ridge National Laboratory was paused in 2020 after attempts to resolve the weld quality issues experienced in 2018-19. Plans have been developed but budget and manpower limitations have prevented further process development and evaluation of welding equipment as recommended in the 2020 milestone report. | Download | 9/17/2021 |
| Microstructure and Mechanical Performance of the Friction Stir Welds Performed on Neutron-Irradiated Steel with Helium, ORNL/TM-2021/2079 | This report describes new experimental results on the microstructure and mechanical performance of the friction stir welds made on neutron-irradiated steel with helium. The report focuses on helium-related issues, specifically, helium-induced degradation in the welded joint, aiming to repair irradiated components of NPPs. | Download | 8/31/2021 |
| Conduct weld campaign (FY-21-1) on irradiated materials provided by the Canadian Nuclear Laboratory (CNL), including baseline post-weld evaluation and testing, ORNL/TM-2020/1673 | A collaborative research on developing advanced welding technologies for irradiated stainless steel 304 materials between CNL and ORNL has been established to support both U.S. and Canadian interests in evaluation of weld repair techniques on irradiated materials to support continuous operation of commercial nuclear power. The work utilizes unique material from the National Research Universal (NRU) reactor and the specialized welding hot cell facility at ORNL. | Download | 8/13/2021 |
| Intermediate-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at the LWR Relevant Temperature, ORNL/TM-2020/1754 | Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. | Download | 9/21/2020 |
| Steam Oxidation of Alloy 718a and Tensile Properties of Select Advanced Replacement Alloys for LWR Core Internals, ORNL/TM-2019/1308 | Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. | Download | 9/12/2019 |
| Recent Technological Advances In Welding Irradiated Austenitic Steel With Helium, ORNL/SPR-2019/1347 | This report describes the post-weld study of weld quality and weld properties of irradiated 304 stainless steel made using both Auxiliary Beam Stress Improved Laser Welding (ABSI-LW) and Friction Stir Welding (FSW) processes and conducted in FY 2019. The results demonstrate that both ABSI-LW and FSW processes, developed in this program, mitigated helium-induced cracking during weld repair of helium-containing irradiated SS304 stainless steels. | Download | 9/2/2019 |
| Develop Parameters and Characterize the Quality of Friction Stir and Laser Weld-Repaired, Irradiated Structural Materials Representative of Extended Reactor Service Life, ORNL/SPR-2019-1035 | This report summarizes the most recent welding campaign on irradiated 304L and 316L at the REDC, and post-welded quality and microstructure characterization, as well, as microhardness testing of ongoing weld campaigns using irradiated stainless-steel alloys at the LAMDA facilities at ORNL. | Download | 4/18/2019 |
| Metallurgical Aspects Influencing the Resistance to Steam Oxidation and Fracture Toughness of Select Advanced Replacement Alloys for LWR Core Internals, ORNL/TM-2018/973 | The corrosion resistance of materials in a water environment is a crucial requirement for core internal materials in light-water reactors (LWRs). High-temperature steam oxidation tests serve as an accelerated life testing method that not only reveals potential failure modes within a short duration but also enables the evaluation of materials’ resistance to accidental scenarios. These tests provide valuable insights into the performance of materials under extreme conditions and contribute to ensuring the reliability and safety of LWR core internal components. | Download | 9/7/2018 |
| Complete Report on Development of Weld Repair Technology M2LW-18OR0406014 | The focus of the research was on post-welding procedures for irradiated alloys, with an emphasis on high helium content stainless steel. The initial measurements of helium content and microstructure characterization indicated successful advanced laser and friction stir welding techniques. The ongoing efforts to weld irradiated alloys with high helium concentrations and thoroughly analyze the results aim to establish validated repair techniques and guidelines for the nuclear industry. | Download | 9/5/2018 |
| Report on the Progress of Weld Development of Irradiated Materials at the Oak Ridge National Laboratory, ORNL/SPR-2018/833 | Report on the Progress of Weld Development of Irradiated Materials at the Oak Ridge National Laboratory, Z. Feng, R. Miller, J. Chen, W. Tang, S. Clark, B. Gibson, M. Vance, G. Frederick, J. Tatman, B. Sutton, April 2018. | Download | 4/18/2018 |
| Proton Irradiation Screening Results of Select Advanced Replacement Alloys for Core Internals | Proton Irradiation Screening Results of Select Advanced Replacement Alloys for Core Internals, G. Was, M. Wang, M. Song, C. Lear, September 2017. | Download | 9/20/2017 |
| Fracture Toughness Evaluation of Select Advanced Replacement Alloys for LWR Core Internals, ORNL/TM-2017/377 | Fracture Toughness Evaluation of Select Advanced Replacement Alloys for LWR Core Internals, ORNL/TM-2017/377, X. Chen, L. Tan, August 25, 2017. | Download | 8/25/2017 |
| Report Summarizing the Status of Second Round Irradiation Experiments and Assessment of Materials Available for Testing Advanced Welding Techniques, M3LW-17OR0406013 | Report Summarizing the Status of Second Round Irradiation Experiments and Assessment of Materials Available for Testing Advanced Welding Techniques, Z. Feng, R. Miller, N. Cetiner, X. Hu, S. Clark, G. Frederick, B. Sutton, July 2017. | Download | 7/18/2017 |
| Report Summarizing the Effort Required to Initiate Welding of Irradiated Materials Within the Welding Cubicle, M3LW-17OR0406012 | Report Summarizing the Effort Required to Initiate Welding of Irradiated Materials Within the Welding Cubicle, Z, Feng, W. Tang, J. Chen, X. Hu, R. Miller, B. Gibson, A. Smith, M. Vance, S. Clark, G. Frederick, J. Tatman, B. Sutton, June 2017. | Download | 6/18/2017 |
| High-Temperature Steam Oxidation Testing of Select Advanced Replacement Alloys for Potential Core Internals, ORNL/TM-2017/228 | High-Temperature Steam Oxidation Testing of Select Advanced Replacement Alloys for Potential Core Internals, ORNL/TM-2017/228, L. Tan, B. Pint, May 19, 2017. | Download | 5/19/2017 |
| Assessment of Radiation Resistance of Selected Alloys in The ARRM Program Using Ion Irradiation, M3LW-16OR0406024 | Assessment of Radiation Resistance of Selected Alloys in The ARRM Program Using Ion Irradiation, G. Was, M. Song, M. Wang, September 30, 2016. | Download | 9/30/2016 |
| Toughness and High-Temperature Steam Oxidation Evaluations of Advanced Alloys for Core Internals, ORNL/TM-2016/371 | Toughness and High-Temperature Steam Oxidation Evaluations of Advanced Alloys for Core Internals, ORNL/TM-2016/371, September 16, 2016. | Download | 9/16/2016 |
| L3 Milestone M3LW-16OR0406015, Provide documentation on the information and states of both the first and second series of neutron irradiate series of B-doped steels for the weld validation testing to begin in FY17 | L3 Milestone M3LW-16OR0406015, Provide documentation on the information and states of both the first and second series of neutron irradiate series of B-doped steels for the weld validation testing to begin in FY17, Z. Feng, N. Cetiner, X. Hu, R. Miller, G. Frederick, B. Sutton, September 2016. | Download | 9/1/2016 |
| Report Detailing Friction Stir Welding Process Development for the Hot Cell Welding System | Report Detailing Friction Stir Welding Process Development for the Hot Cell Welding System, M3LW-16OR0406014, W. Tang, B. Gibson, Z. Feng, S. Clark, A. Peterson, G. Frederick, September 2016. | Download | 9/1/2016 |
| Report on the Installation of the Integrated Welding Hot Cell at ORNL Building 7930, M2LW-16OR0406013 | Report on the Installation of the Integrated Welding Hot Cell at ORNL Building 7930, M2LW-16OR0406013, Z. Feng, W. Tang, R. Miller, B. Gibson, A. Peterson, J. Tatman, G. Frederick, June 30, 2016. | Download | 6/30/2016 |
| Status update of advanced alloys for Advanced Radiation Resistant Materials Program | Status update of advanced alloys for Advanced Radiation Resistant Materials Program, L. Tan, D. Hoelzer, and J. Busby, Oak Ridge National Laboratory, March 2014. | Download | 3/1/2014 |
| Use of Computational Model to Design and Optimize Welding Conditions to Suppress Helium Cracking during Welding, W. Zhang Z. Feng and E. Willis, June 2012. | Use of Computational Model to Design and Optimize Welding Conditions to Suppress Helium Cracking during Welding, W. Zhang Z. Feng and E. Willis, June 2012. | Download | 6/1/2012 |
| Develop Baseline Computational Model for Proactive Welding Stress Management to Suppress Helium-Induced Cracking During Weld Repair, W. Zhang and Z. Feng, ORNL, September 2011. | Develop Baseline Computational Model for Proactive Welding Stress Management to Suppress Helium-Induced Cracking During Weld Repair, W. Zhang and Z. Feng, ORNL, September 2011. | Download | 9/1/2011 |
| Complete the first phase of the comprehensive characterization of repair welding performed on irradiated Ni alloy182 using stress improved laser welding in collaboration with EPRI,ORNL/TM-2023/3109 | This report describes the research activities conducted in FY 2023 on post-weld evaluation and characterization of the quality and properties of welds made on irradiated Ni-base alloy 182 (with boron concentration up to 15 wppm) by the advanced laser repair welding technology developed under the DOE Light Water Reactor Sustainability Program and EPRI LTO program. The research represents a major progress in repair welding of highly irradiated helium containing reactor internals and its feasibility to use the ABSI-LW technology developed in this program. | Download | 09/15/2023 |
| Report Title | Brief Narrative | Link | Date |
|---|---|---|---|
| Spread Spectrum Time Domain Reflectometry (SSTDR) and Frequency Domain Reflectometry (FDR) for Detection of Cable Anomalies Using Machine Learning, PNNL-34821 | Cables are initially qualified for nuclear power plant use for 40 years. As plants extend their operating license to 60 and 80 years, continued use of these cables must shift to a performance-based approach since it is cost prohibitive to completely replace cables that are Frequency domain reflectometry (FDR) is one of these test methods that is being more broadly accepted and used because it not only detects anomalies along the cable with a low-voltage signal that does not stress the cable insulation, but the technique also locates the anomalies. This supports follow-up local inspection and local repair or partial replacement of a damaged cable segment. Currently, FDR testing is only applied to cables that are taken out of service since the test instrument would be damaged by operational voltages. A related technology that has found some acceptance in the aircraft and rail industry is spread spectrum time domain reflectometry (SSTDR). This technology has been implemented with a custom commercial instrument by LiveWire Innovation Inc. that is designed to operate on live cables up to 1000 volts.likely still capable of performing their design function. A variety of cable tests are available and are commonly applied during outages when the cables can be taken out of service. | Download | 9/26/2023 |
| Extended Bandwidth Spread Spectrum Time Domain Reflectometry Cable Test for Thermal Aging, Low Resistance Fault, and Water Detection, PNNL-34815 | In 2022, researchers at Pacific Northwest National Laboratory (PNNL) used the Accelerated and Real-Time Environmental Nodal Assessment (ARENA) cable and motor test bed to characterize spread spectrum time domain reflectometry (SSTDR) and compare the responses of an SSTDR instrument to those of a frequency domain reflectometry (FDR) instrument. Results showed both techniques could detect and locate cable anomalies such as phase-to-phase low resistance and shorts, thermal insulation damage, mechanical insulation damage, and the presence or absence of water in some conditions. The SSTDR tests used a commercial instrument provided by LiveWire Innovations Inc. This commercial instrument performed tests at 6, 12, 24, and 48 MHz bandwidth. The results of these tests were compared to FDR tests where bandwidths could be extended up to 1.3 GHz, although the best responses for cable tests were from 100 to 500 MHz. Lower bandwidth signals can propagate better along the cable while higher bandwidths have higher resolution for impedance change reflections allowing more precise indication of location and separation of anomalies. | Download | 9/26/2023 |
| The Mechanism of Irradiation Assisted Stress Corrosion Cracks in Stainless Steels | The objective of this Light Water Reactor Sustainability (LWRS) project is to determine the mechanism of irradiation-assisted stress corrosion cracking (IASCC) in austenitic stainless steels in a PWR primary water environment and to propose mitigation strategies. A novel miniaturized four-point bend test was used to determine the crack initiation stress and to relate it to the microstructure features responsible for crack initiation. In this current work, we lay out the mechanism of IASCC as deduced from work in this LWRS program, complimentary programs and work done by others over the past 60 years. Despite evidence of this degradation mode that dates back to the 1960s, the mechanism by which it occurs has remained elusive. | Download | 9/26/2023 |
| Complete the additional microstructural evaluation and SCC CGR testing on two heats of aged Alloy 152, ANL/LWRS-23/1 | Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is critical for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. | Download | 9/26/2023 |
| Applying grain-boundary sensitive electrochemical scanning probe techniques to evaluate intergranular degradation of irradiated and deformed stainless steels | : In nuclear power plants, irradiation assisted stress corrosion cracking (IASCC) of the critical structural components can cause frequent shutdown of the nuclear reactor. This results in the power generation loss and incurring high maintenance cost. IASCC is a complex problem where the synergetic effect of irradiation, mechanical load, and corrosion activity all comes into play, thereby making its mechanism difficult to understand. Irradiation assisted elemental segregation (e.g., Cr-depletion at grain boundaries) can be the major reasons to induce IASCC, but the corrosion activity related to such compositional heterogeneities has not been fully understood. Thus, in present study, we address the effect of ion irradiation on the electrochemical corrosion over the localized features such as grain interiors, grain boundaries, and dislocation channels. | Download | 9/26/2023 |
| FY 2023 Progress on Stress Corrosion Crack Testing of Ni-Base Alloys in PWR Primary Water | The first part of this report documents the final year progress of a three-year research effort on evaluating the SCC initiation and growth behavior of Ni-base alloys in LiOH vs. KOH-containing PWR primary water. | Download | 8/29/2023 |
| Preparation for Stress Corrosion Crack Initiation Testing of Austenitic Stainless Steels in PWR Primary Water | Operating experience of Type 304L 316 L austenitic stainless steels SS in PWR primary circuits have generally been excellent, but increasing intergranular stress corrosion cracking (IGSCC) incidents have been reported in free-flowing PWR primary water in recent years (e.g., Japan Ohi-3: SCC in 316SS HAZ in Pressurizer Spray Line and Multiple French N4 and P4 units: SCC in Safety Injection Lines and Residual Heat Removal Lines), posing a potentially serious emerging issue affecting nuclear power plants availability. A better understanding of the IGSCC initiation mechanism and its threats to plants are required to inform utility and regulatory on a proactive management strategy. In preparation for devising a detailed testing plan to address this need, this report reviews available field experience and laboratory studies on SCC initiation of austenitic SS in normal PWR primary water environments. Knowledge and technical gaps are identified, and a near-term action plan is proposed. | Download | 6/26/2023 |
| Microstructural characterizations of the second high fluence baffle-former bolt retrieved from a Westinghouse two-loop downflow type PWR, ORNL/TM-2022/2668 | As one of the PWR internal components, BFBs are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts. In support of evaluating long-term operational performance of materials used in core internal components, the ORNL, through the DOE, LWRS Program, MRP has harvested two high fluence BFBs from a commercial Westinghouse two-loop downflow type PWR. | Download | 10/10/2022 |
| The role of grain boundary oxides in the susceptibility to irradiation assisted stress corrosion cracking for high dose 304 stainless steel under pressurized water reactor relevant conditions, M3LW-22OR0402028 | The objective of this report is to pin down the mechanism of IASCCi n high dose (neutron-irradiated) solution-annealed 304 stainless steel in PWR primary water environment and to propose mitigation strategies. A novel miniaturized four-point bend test was used to determine the crack initiation stress and to relate it to the microstructure features responsible for crack initiation. In this current work, our efforts were devoted to identifying the role of GB oxidation in IASCC. | Download | 9/26/2022 |
| Electrochemical probing of microstructural heterogeneities in irradiated and deformed stainless steel | IASCC in nuclear components amounts to higher energy production cost and reduction in productivity of the reactor during the repair period. IASCC is a multifaceted problem comprising various phenomena operating simultaneously, making it challenging to exactly describe the full spectra of the phenomenon mechanistically. Influence of the irradiation is attributed to the acceleration of SCC phenomenon with changes to both material properties and water chemistry due to irradiation contributing to this acceleration. | Download | 9/18/2022 |
| Hybrid Artificial Intelligence-Machine Learning and Finite Element-based Digital Twin Predictive Modeling Framework for PWR Coolant System Components: Updates on Multi-Time-Series-3D-Location Dependent Usage Factor Prediction, ANL/LWRS-22/1 | The LTO of NPP beyond their original design life of 40 years can lead to more material damage associated with cyclic fatigue under thermal-mechanical loading cycles and associated long-term exposure of reactor material to the deleterious reactor-coolant environments. However, under this LTO condition, the reactor components can still safely operate but may require more frequent NDE of reactor components. Requiring frequent NDE inspections may lead to frequent NPP shutdowns which can lead to power outages and additional NDE inspection cost-related economic loss. | Download | 9/17/2022 |
| Effect of thermal aging on microstructure and crack growth response of Alloy 152 Weld, ANL/LWRS-22/2 | Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for PWR vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily SCC. | Download | 9/10/2022 |
| Complete the stress corrosion crack initiation and crack growth response of Ni-base alloys in KOH vs. LiOH PWR primary water chemistry, M3LW-22OR0402033 | The U.S. nuclear industry is considering replacing lithium hydroxide (LiOH) with potassium hydroxide (KOH) for pH control in pressurized water reactor (PWR) primary water for economic reasons. Among the many aspects of reactor operation that need to be assessed before switching to KOH, it is necessary to evaluate the stress corrosion cracking (SCC) response of Ni-base alloys in a KOH environment. | Download | 7/29/2022 |
| Stress Corrosion Cracking of Ni-base Alloys in PWR Primary Water Containing KOH vs. LiOH, M3LW-22OR0402033 | The U.S. nuclear industry is considering replacing LiOH with potassium hydroxide (KOH) for pH control in PWR primary water for economic reasons. | Download | 7/11/2022 |
| Toward an understanding of straining mode, grain boundary oxidation and localized deformation on intergranular cracking of neutron irradiated austenitic stainless steels in pressurized water reactor relevant conditions, M2LW-21OR0402023 | The main objective of the report is to identify the mechanism of IASCC in highly irradiated solution-annealed 304 and cold-worked 316 stainless steels in PWR primary water environment and to recommend mitigation strategies. The four-point bend test was used to determine the crack initiation stress and then, to identify the microstructure features responsible for IASCC initiation. | Download | 9/27/2021 |
| Fracture Toughness and Fatigue Crack Growth Rate Testing of Baffle-Former Bolts Harvested from a Westinghouse Two-Loop Downflow Type PWR, ORNL/TM-2021/2264 | As one of the PWR internal components, baffle-former bolts (BFBs) are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts. | Download | 9/22/2021 |
| Evaluation of Stress Corrosion Cracking Behavior of Ni-base Alloys in PWR Primary Water Containing KOH vs. LiOH, M2LW-21OR0402036 | The U.S. nuclear industry is considering replacing LiOH with potassium hydroxide (KOH) for pH control in PWR primary water for economic reasons. | Download | 9/21/2021 |
| Electrochemical profiling of physical damage in nuclear core components, M2LW-21OR0402027 | Alloy degradation has been accompanied by our metal-based civilization for thousands of years; its costs are still beyond estimation as of today. After entering the nuclear-era, metals have been acclimatized to more acute environments in NPPs and nuclear waste disposal facilities, wherein alloys can be damaged physically (e.g., via deformation and irradiation) electrochemically, or via synergetic pathways. | Download | 9/20/2021 |
| Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690, ANL/LWRS-21/1 | Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for PWR vessels, because of their increased resistance to SCC relative to Alloys 600, 82, and 182. Many of these reactors are expected to operate for 40-80 years. | Download | 9/9/2021 |
| Evaluation of Critical Parameters to Model Stress Corrosion Crack Initiation in Alloy 600 and Alloy 182 in PWR Primary Water, M2LW-20OR0402036 | Conventional Ni-base Alloy 600 is still in use in many operating domestic pressurized water reactors (PWR), especially in thick-section components where primary water stress corrosion crack (PWSCC) has been observed and is a concern to PWR life extension. While SCC initiation is considered to encompass the majority of the lifespan of Alloy 600 components, mechanistic understanding of this process remained insufficient for better lifetime predictions, safety assessments and risk management during extended operation of nation’s existing LWR fleet. | Download | 9/21/2020 |
| Evaluation of the stress and fluence dependence of irradiation assisted stress corrosion crack initiation in high fluence austenitic stainless steels under pressurized water reactor relevant conditions, M2LW-20OR0402023 | The program consists of two major tasks; determination of the irradiation assisted stress corrosion cracking (IASCC) susceptibility of solution annealed (SA) 304 and cold-worked (CW) 316 stainless steel (SS) in PWR primary water environment. This was done using two principle test methods; the constant extension rate tests (CERT) to determine the stress-strain and fracture behaviors using tensile bars. | Download | 9/21/2020 |
| A Hybrid AI/ML and Computational Mechanics Based Approach for Time-Series State and Fatigue Life Estimation of Nuclear Reactor Components, ANL/LWRS-20/01 | Environmental fatigue modeling is a complex problem due to multiple failure modes and their intermixing. The failure modes are function of various underlying causes in addition to the corrosive effect of reactor coolant environment. Some of the major causes are time-dependence of material associated with cyclic loading, load sequence effect associated with random/variable amplitude loading, effect of strain amplitude and rates, effect of varying temperature (along both temporal and spatial directions) and the effect of mean strain and stress. | Download | 9/21/2020 |
| Analysis of Deformation Localization Mechanisms in Highly Irradiated Austenitic Stainless Steel via In Situ Techniques, ORNL/TM-2020/1739 | This report describes new experimental results on deformation localization phenomena in austenitic steels irradiated to high-damage doses. Advanced in situ tests were performed with specimens irradiated up to 95 dpa at ORNL’s Low Activation Materials Development and Analysis facility. The tensile tests inside an SEM allowed for EBSD analysis and studying cracking and fracture initiation processes. Experiments also employed high-resolution EBSD to record and analyze Kikuchi patterns. | Download | 9/21/2020 |
| Quantifying micro-galvanic corrosion in stainless steels activated by post-yielding microstructures | Plastic strain renders nuclear reactor components more susceptible to SCC and premature failure. In this study, for the first time, the strain activation of micro-galvanic corrosion in stainless steels was identified as a cause for SCC. Corrosion and surface reactivity of progressively elongated steels were analyzed using multimodal and multiscale approaches. | Download | 9/21/2020 |
| Long-Term Crack Initiation Behavior of Alloy 690 and Its Weld Metals in PWR Primary Water | This report updates the ongoing long-term constant load SCC initiation testing on 42 Alloy 690 specimens from seven control rod drive mechanism/plate/bar heats with different CW levels and thermal-mechanical histories where susceptibility to SCC growth has been documented in tests at PNNL. | Download | 3/10/2020 |
| A System-Level Framework for Fatigue Life Prediction of a PWR Pressurizer-Surge-Line Nozzle under Design-basis Loading, ANL/LWRS-19/01 | This report presents a comprehensive material properties database, formatted in an industry-standard SQL format, focusing on key materials found in a reactor coolant system. These materials include 316 stainless steel (SS), 508 LAS, SMWs between 316 SS, and DMWs such as 508LAS and 316SS. The DMWs encompass In-82 filler welds and In-182 butter welds. The material properties included in the database are derived from 21 tensile tests conducted under the LWRS program, with the tests taking place either during FY 2019 or earlier. | Download | 9/5/2019 |
| Elucidating The Grain-Orientation Dependent Oxidation Rates Of Austenitic Stainless Steels, M2LW-19OR0402028 | This study focuses on the behavior of austenitic stainless steel, which is commonly used in critical core-internal components of nuclear power plants. The complex microstructure of stainless steel, characterized by multiple grains with diverse orientations, has been found to exhibit grain orientation-dependent oxidation rates in nuclear reactor environments. To gain a better understanding of this phenomenon, the researchers investigated the oxidation behaviors of 316L stainless steel using vertical scanning interferometry (VSI). | Download | 9/3/2019 |
| Grain Boundary Microstructure Effects On Stress Corrosion Crack Initiation Mechanisms In Alloy 600 And Alloy 690 | Grain Boundary Microstructure Effects On Stress Corrosion Crack Initiation Mechanisms In Alloy 600 And Alloy 690, Z. Zhai, M. Toloczko, M. Olszta, S. Bruemmer, September 2019. | Download | 9/3/2019 |
| Irradiation Assisted Stress Corrosion Cracking of Highly Irradiated 300-Series Stainless Steels in PWR Primary Water Environment. M2LW-19OR0402026 | The program consists of two major tasks; determination of the irradiation assisted stress corrosion cracking (IASCC) susceptibility of highly irradiated 304 and cold-worked 316 stainless steel in PWR primary water environment. This was done using two principle test methods; the constant extension rate test (CERT) to determine the stress-strain behavior and the fracture behavior using tensile bars, and a four-point bend test to determine the stress to crack initiation and the microstructural features responsible for it. | Download | 9/1/2019 |
| Analysis of Localized Deformation Processes in Highly Irradiated Austenitic Stainless Steel through In Situ Techniques, ORNL/TM-2019/1274 | This report describes new experimental results obtained during in situ mechanical tests with neutron irradiated (10.7 dpa) austenitic 304L steel specimens. The in situ tensile tests inside a scanning electron microscope (SEM) was accompanied by modern high-resolution electron backscatter diffraction analysis. The HR-EBSD approach allowed for investigating and quantifying internal stresses and dislocation structures in nondeformed and deformed steels, materials used in light water reactor internal components. | Download | 8/9/2019 |
| Post-Irradiation Examination of High Fluence Baffle-Former Bolts Retrieved from a Westinghouse Two-Loop Downflow Type PWR, ORNL/TM-2019/1251 | This report presents the findings of a post-irradiation examination conducted on specific specimens from two high fluence BFB systems. The examination involved the use of electron microscopy and microhardness characterization techniques. The study reveals that the radiation-induced defects observed in the material of interest are similar to the defects induced by neutron radiation in stainl | Download | 7/31/2019 |
| Material Condition Effects on Stress Corrosion Crack Initiation Of Cold-Worked Alloy 600 In Pwr Primary Water Environments, M3LW-19OR0402037 | This research project aims to address the understanding of crack initiation, which is one of the least known aspects of SCC in pressure boundary components of LWRs. The focus is on investigating the influence of various material factors such as composition, processing, microstructure, and strength, as well as environmental factors including temperature, water chemistry, electrochemical potential, and stress, on the susceptibility of corrosion-resistant, nickel-base alloys to SCC. | Download | 4/1/2019 |
| Implementation of ANL’s Mechanics Based Evolutionary Fatigue Modeling Through ABAQUSWARP3D Based High-Performance Computing Framework, ANL/LWRS-18/01 | These findings demonstrate the potential for a comprehensive mechanistic approach, rather than relying solely on S-N curves, to evaluate the fatigue life of safety-critical components in nuclear reactors. This approach holds promise not only for nuclear reactor components but also for other safety-critical applications such as aircraft and aero-engines. | Download | 10/18/2018 |
| Fracture Resistance of Cast Stainless Steels after Thermal Aging for up to 10000 Hours, PNNL-27923 | Based on the initial findings of this research, it can be concluded that aged CASS materials with δ-ferrite contents below approximately 20% are unlikely to experience a substantial reduction in static fracture toughness or embrittlement throughout the extended lifetimes of reactors. | Download | 9/20/2018 |
| Stress Corrosion Crack Initiation and Propagation under Different Corrosion Environments and the Role of Post-Irradiation Annealing on Irradiation Assisted Stress Corrosion Crack Mitigation, M2LW-18OR0402026 | This research aims to enhance our understanding of the parameters and microstructural characteristics that govern IASCC, develop a post-irradiation annealing process to mitigate IASCC, and establish a basis for constructing predictive models for IASCC. Ultimately, this program seeks to provide a comprehensive understanding of IASCC and contribute to the development of effective strategies for its prevention and mitigation. | Download | 9/20/2018 |
| Thermodynamic and kinetic model of phase stability in austenitic steel under light water reactor conditions | Austenitic stainless steels and low-alloy ferritic steels play vital roles in various applications within the nuclear industry. Austenitic stainless steels, renowned for their exceptional strength, corrosion resistance, and formability, are extensively used as structural materials in nuclear reactor cores. Nevertheless, the harsh conditions of the reactor core, characterized by elevated temperatures and neutron irradiation, pose challenges such as materials degradation, including void swelling and precipitation hardening. | Download | 8/31/2018 |
| Revealing how alkali cations affect stainless steel passivation in alkaline aqueous environments, UCLA-LWRS-2018-1 | Stainless steel is a ubiquitous material that finds use in structural and core-internal components in nuclear power plants. Stainless steel features superior corrosion resistance due to the formation of passivating iron and/or chromium oxides on its surfaces. | Download | 8/18/2018 |
| Preparation and Analysis of Austenitic Stainless Steel Samples Irradiated at Very High Damage Doses, ORNL/TM-2018/113792 | Preparation and Analysis of Austenitic Stainless Steel Samples Irradiated at Very High Damage Doses, ORNL/TM-2018/113792, M. Gussev, S. Clark, J. Dixon, K. Leonard, July 2018. | Download | 7/28/2018 |
| Stress Corrosion Crack Initiation Mechanisms of Nickel-Base Alloys in Simulated PWR Primary Water, M2LW-18OR0402026 | Stress Corrosion Crack Initiation Mechanisms of Nickel-Base Alloys in Simulated PWR Primary Water, M3LW-18OR0402033, Z. Zhai, M. Toloczko, M. Olszta, K. Kruska, D. Schreiber, S. Bruemmer, May 2018. | Download | 5/28/2018 |
| PNNL Presentations on SCC Initiation at the 2017 EPRI Alloy 690/152/52 Research Collaboration Meeting Milestone Report: M3LW-18OR0402032 | PNNL Presentations on SCC Initiation at the 2017 EPRI Alloy 690/152/52 Research Collaboration Meeting Milestone Report: M3LW-18OR0402032, S. Bruemmer,, Z. Zhai, M. Toloczko, November 28, 2017. | Download | 11/18/2017 |
| Effect of Thermal Aging in Stainless Steel Welds: 2nd Year Progress Report of I-NERI Collaboration, PNNL-27013 | Effect of Thermal Aging in Stainless Steel Welds: 2nd Year Progress Report of I-NERI Collaboration, PNNL-27013, T. Byun, T. Lach, D. Collins, C. Jang, October 2017. | Download | 10/31/2017 |
| Final Report on CFD and Thermal-Mechanical Stress Analysis of PWR Surge Line under Transient Condition Thermal Stratification and an Evolutionary Cyclic Plasticity Based Transformative Fatigue, ANL/LWRS-17/03 | Final Report on CFD and Thermal-Mechanical Stress Analysis of PWR Surge Line under Transient Condition Thermal Stratification and an Evolutionary Cyclic Plasticity Based Transformative Fatigue Evaluation Approach without Using S~N Curve, ANL/LWRS-17/03, S. Mohanty, B. Barua, J. Listwan, S. Majumdar, K. Natesan, September 2017. | Download | 9/20/2017 |
| Localized Deformation Investigation In Irradiated Materials Via Electron Microscopy And In Situ Testing, ORNL/TM-2017/507 | Localized Deformation Investigation In Irradiated Materials Via Electron Microscopy And In Situ Testing, ORNL/TM-2017/507, M. Gussev, P. Edmondson, G. de Bellefon, B. Eckhart, J. Dixon, K. Leonard, September 2017. | Download | 9/20/2017 |
| Stress Corrosion Crack Initiation and Propagation under Different Corrosion Environmental conditions, and the Role of Post-Irradiation Annealing on Irradiation Assisted Stress Corrosion Cracking | Stress Corrosion Crack Initiation and Propagation under Different Corrosion Environmental conditions, and the Role of Post-Irradiation Annealing on Irradiation Assisted Stress Corrosion Cracking, G. Was, W. Kuang, R. Bhambroo, September 2017. | Download | 9/20/2017 |
| Stress Corrosion Crack Intiation of Alloy 600 in Simulated PWR Primary Water M2LW-17OR0402034 | Stress Corrosion Crack Intiation of Alloy 600 in Simulated PWR Primary Water, M2LW-17OR0402034, Z. Zhai, M. Toloczko, M. Olszta, D. Schreiber, S. Bruemmer, September 2017. | Download | 9/20/2017 |
| Development of Computational Tools for Modeling Thermal and Radiation Effects on Grain Boundary Segregation and Precipitation in Fe-Cr-Ni-based Alloys, ORNL/TM-2017/438 | Development of Computational Tools for Modeling Thermal and Radiation Effects on Grain Boundary Segregation and Precipitation in Fe-Cr-Ni-based Alloys, ORNL/TM-2017/438, Y. Yang, August 2017. | Download | 8/18/2017 |
| Mechanical and Microstructural Characteristics of Cast Stainless Steels after Thermal Aging for 10000 Hours, PNNL-26656 | Mechanical and Microstructural Characteristics of Cast Stainless Steels after Thermal Aging for 10000 Hours, PNNL-26656, T. Byun, T. Lach, D. Collins, E. Barkley, F. Yu, August 2017. | Download | 8/18/2017 |
| Predictive model for swelling accumulation in austenitic steels under light water reactor relevant conditions, ORNL/LTR-2017/407 | Predictive model for swelling accumulation in austenitic steels under light water reactor relevant conditions, ORNL/LTR-2017/407, S. Golubov, A. Barashev, August 2017. | Download | 8/18/2017 |
| Effect of Swelling on Irradiation-Assisted Stress Corrosion Cracking, INL/EXT-17-42863 | Effect of Swelling on Irradiation-Assisted Stress Corrosion Cracking, INL/EXT-17-42863, S. Teysseyre, Juy 2017. | Download | 7/29/2017 |
| PNNL Presentation on SCC Initiation at the 2017 International Cooperative Group Meeting on Environment-Assisted Cracking, | PNNL Presentation on SCC Initiation at the 2017 International Cooperative Group Meeting on Environment-Assisted Cracking, S. Bruemmer, Z. Zhai, M. Toloczko, May 7, 2017. | Download | 5/7/2017 |
| 3D-FE Modeling of 316 SS under Strain-Controlled Fatigue Loading and CFD Simulation of PWR Surge Line, ANL/LWRS-17/01 | 3D-FE Modeling of 316 SS under Strain-Controlled Fatigue Loading and CFD Simulation of PWR Surge Line, ANL/LWRS-17/01, S. Mohanty, B. Barua, J. Listwan, S. Majumdar, K. Natesan, March 2017. | Download | 3/1/2017 |
| Development and Initial Test Results of the In-Situ Characterization of Irradiated Materials under Deformation, ONRL/TM-2017/111 | Development and Initial Test Results of the In-Situ Characterization of Irradiated Materials under Deformation, ONRL/TM-2017/111, M. Gussev, P. Edmondson, B. Eckhart, K. Leonard, February 2017. | Download | 2/14/2017 |
| SSC Initiation in Alloy 600 and Alloy 690, M3LW-17OR0 | SSC Initiation in Alloy 600 and Alloy 690, Milestone: M3LW-17OR0, S. Bruemmer, Z. Zhai, M. Toloczko, December 19, 2016. | Download | 12/19/2016 |
| Incorporation of copper-rich precipitation model into developed Ni-Mn-Si precipitate development models | Incorporation of copper-rich precipitation model into developed Ni-Mn-Si precipitate development models September 30, 2016 Milestone, M. Mamivand, H. Ke, S. Shu, D. Morgan, G. Odette, P. Wells, N. Almirall, September 30, 2016. | Download | 9/30/2016 |
| Study of High Fluence Radiation-induced Swelling and Hardening under Light Water Reactor Conditions, ORNL/LTR-2016/549 | Study of High Fluence Radiation-induced Swelling and Hardening under Light Water Reactor Conditions, ORNL/LTR-2016/549, S. Golubov, A. Barashev, R. Stoller, September 2016. | Download | 9/1/2016 |
| Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components, ANL/LWRS-16/03 | Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components, ANL/LWRS-16/03, S. Mohanty, B. Barua, W. Soppet, S. Majumdar, K. Natesan, September 2016. | Download | 9/1/2016 |
| Precursor Damage Evolution and Stress Corrosion Crack Initiation of Cold-worked Alloy 690 in PWR Primary Water,M2LW-16OR0402034 | Precursor Damage Evolution and Stress Corrosion Crack Initiation of Cold-worked Alloy 690 in PWR Primary Water, M2LW-16OR0402034, Z. Zhai, M. Toloczko, K. Kruska, N. Overman, M. Olszta, S. Bruemmer, September 2016. | Download | 9/1/2016 |
| Development of Computational Tools for Predicting Thermal- and Radiation-Induced Solute Segregation at Grain Boundaries in Fe-based Alloys, ORNL/TM-2016/460 | Development of Computational Tools for Predicting Thermal- and Radiation-Induced Solute Segregation at Grain Boundaries in Fe-based Alloys, ORNL/TM-2016/460, Y. Yang, September 2016. | Download | 9/1/2016 |
| Mechanical Properties of 304L and 316L Austenitic Stainless Steels after Thermal Aging for 1500 Hours, PNNL-25854 | Mechanical Properties of 304L and 316L Austenitic Stainless Steels after Thermal Aging for 1500 Hours, PNNL-25854, T. Byun, T. Lach, September 2016. | Download | 9/1/2016 |
| Mechanical Properties of Model Cast Austenitic Stainless Steels after Thermal Aging for 1500 Hours, PNNL-25377 | Mechanical Properties of Model Cast Austenitic Stainless Steels after Thermal Aging for 1500 Hours, PNNL-25377, T. Sang Byun, N. Overman, T. Lach, April 2016. | Download | 4/1/2016 |
| Effect Of Thermo-Mechanical Processing And Aging On The Stress Corrosion Cracking Resistance Of Alloy 690 | Effect Of Thermo-Mechanical Processing And Aging On The Stress Corrosion Cracking Resistance Of Alloy 690, 2011-01-K, March 2016. | Download | 3/31/2016 |
| Finite Element Analysis Based Thermal-Mechanical Stress Analysis of PWR Pressure Vessel with/without Pre-existing Crack to Study the Effect of Cyclic Hardening Material Properties Under Grid Load Following Mode: An Interim Reportt, ANL/LWRS-16/01 | Finite Element Analysis Based Thermal-Mechanical Stress Analysis of PWR Pressure Vessel with/without Pre-existing Crack to Study the Effect of Cyclic Hardening Material Properties Under Grid Load Following Mode: An Interim Report, ANL/LWRS-16/01, S. Mohanty, W. Soppet, S. Majumdar, K. Natesan, March 2016. | Download | 3/1/2016 |
| Thermodynamic and Kinetic Model Development of Thermal Segregation of Phosphorus in Iron, Fe-Cr and Fe-Ni Systems – Towards the Development of a Comprehensive Solute Segregation Model, ORNL/TM-2016/93 | Thermodynamic and Kinetic Model Development of Thermal Segregation of Phosphorus in Iron, Fe-Cr and Fe-Ni Systems – Towards the Development of a Comprehensive Solute Segregation Model, ORNL/TM-2016/93, Y. Yang, March 2016. | Download | 3/1/2016 |
| Modeling Strategy to Assess Radiation Induced Segregation and Phase Stability in Austenitic Steels in Light Water Reactors During Extended service: Milestone 7 Report | Modeling Strategy to Assess Radiation Induced Segregation and Phase Stability in Austenitic Steels in Light Water Reactors During Extended Service: Milestone 7 Report, M. Mamivand, Y. Yang, D. Morgan, September 2015. | Download | 9/30/2015 |
| Summary of Stress Corrosion Crack Initiation Measurements and Analyses on Alloy 600 and Alloy 690 | Summary of Stress Corrosion Crack Initiation Measurements and Analyses on Alloy 600 and Alloy 690, Z. Zhai, M. J. Olszta, M. B. Toloczko and S. M. Bruemmer, September 2015. | Download | 9/1/2015 |
| Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions ANL/LWRS-15/02 | Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions, S. Mohanty, W. Soppet, S. Majumdar, and K. Natesan, ANL/LWRS-15/02 September 2015. | Download | 9/1/2015 |
| Development of a robust modeling tool for radiation-induced segregation in austenitic stainless steels ORNL/TM-2015/479 | Development of a robust modeling tool for radiation-induced segregation in austenitic stainless steels, Y. Yang, K. Field, T. Allen, J. T. Busby, ORNL/TM-2015/479, September 2015. | Download | 9/1/2015 |
| System-Level Heat Transfer Analysis, Thermal-Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor: A Preliminary Study, ANL/LWRS-15/01 | System-Level Heat Transfer Analysis, Thermal-Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor: A Preliminary Study, S. Mohanty, W. Soppet, S. Majumdar, and K. Natesan, ANL/LWRS-15/01, Argonne National Laboratory, April 2015. | Download | 5/1/2015 |
| Assessment of Cable Aging Equipment, Status of Acquired Materials, and Experimental Matrix, PNNL-24198 | Assessment of Cable Aging Equipment, Status of Acquired Materials, and Experimental Matrix at the Pacific Northwest National Laboratory, L. Fifield, M. Westman, A. Zwoster, and B. Schwenzer, PNNL-24198, Pacific Northwest National Laboratory, March 2015. | Download | 3/1/2015 |
| Reestablishing Capability at PNNL and Aging and Testing Schedule | Reestablishing Capability at PNNL and Aging and Testing Schedule, T. Byun, Pacific Northwest National Laboratory, January 2015. | Download | 1/1/2015 |
| Environmental Effect of Evolutionary Cyclic Plasticity Material Parameters of 316 Stainless Steel: An Experimental & Material Modeling Approach, ANL/LWRS-14/01 | Environmental Effect of Evolutionary Cyclic Plasticity Material Parameters of 316 Stainless Steel: An Experimental & Material Modeling Approach, S. Mohanty, W. K. Soppet, S. Majumdar, and K. Natesan, Argonne National Laboratory, ANL/LWRS-14/01, September 2014. | Download | 9/1/2014 |
| Baseline Characterization of Cast Stainless Steels, ORNL/TM-2014/446 | Baseline Characterization of Cast Stainless Steels, T.S. Byun and Y. Yang, Oak Ridge National Laboratory, ORNL/TM-2014/446, September 2014. | Download | 9/1/2014 |
| A Model of Radiation-Induced Microstructural Evolution, ORNL/LTR-2014/487 | A Model of Radiation-Induced Microstructural Evolution, A. Barashev, S. Golubov, and R. Stoller, ORNL/LTR-2014/487, September 2014. | Download | 9/1/2014 |
| Stress Corrosion Crack Initiation of Cold-Worked Alloy 600 and Alloy 690 in PWR Primary Water | Stress Corrosion Crack Initiation of Cold-Worked Alloy 600 and Alloy 690 in PWR Primary Water, S. Bruemmer, M. Olszta, D. Schreiber, and M. Toloczko, Pacific Northwest National Laboratory, September 2014. | Download | 9/1/2014 |
| Analysis of Phase Transformation Studies in Solute Addition Alloys, ORNL/TM-2014/303 | Analysis of Phase Transformation Studies in Solute Addition Alloys, L. Tna, K. Field, and J. Busby, Oak Ridge National Laboratory, ORNL/TM-2014/303, August 2014. | Download | 8/1/2014 |
| BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates, INL/EXT-14-31703 | BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates, S. Teysseyre, Idaho National Laboratory, INL/EXT-14-31703, April 2014. | Download | 4/1/2014 |
| Corrosion and Stress Corrosion Crack Initiation of Cold-Worked Alloy 690 in PWR Primary Water | Corrosion and Stress Corrosion Crack Initiation of Cold-Worked Alloy 690 in PWR Primary Water, T.L. Dickson, M.B. Toloczko, M.J. Olszta, D.K. Schreiber, and S.M. Bruemmer, PNNL on behalf of ORNL, September 2013. | Download | 9/1/2013 |
| Analysis of Deformation Mode Changes in Irradiated Materials using Bend Tests and Finite Element Modeling, ORNL/TM-2013/217 | Analysis of Deformation Mode Changes in Irradiated Materials using Bend Tests and Finite Element Modeling, M. N. Gussev, J. T. Busby, ORNL/TM-2013/217, June 2013. | Download | 6/1/2013 |
| Modeling Strategy to Assess Radiation Induced Segregation and Phase Stability in Austenitic Steels in Light Water Reactors During Extended Service | Modeling Strategy to Assess Radiation Induced Segregation and Phase Stability in Austenitic Steels in Light Water Reactors During Extended Service, L. Barnard, D. Morgan, B. Wirth, June 2013. | Download | 6/1/2013 |
| Preliminary Report on Assessment of Environmentally-Assisted Fatigue for LWR Extended Service Conditions, S. Mohanty, W. K. Soppet, S. Majumdar, K. Natesan, ANL, March 2013 | Preliminary Report on Assessment of Environmentally-Assisted Fatigue for LWR Extended Service Conditions, S. Mohanty, W. K. Soppet, S. Majumdar, K. Natesan, ANL, March 2013. | Download | 3/1/2013 |
| Microstructure, Corrosion and Stress Corrosion Crack Initiation of Alloy 600 in PWR Primary Water Environments, M. J. Olszta, D. K. Schreiber, M. B. Toloczko, S. M. Bruemmer, PNNL, March 2013 | Microstructure, Corrosion and Stress Corrosion Crack Initiation of Alloy 600 in PWR Primary Water Environments, M. J. Olszta, D. K. Schreiber, M. B. Toloczko, S. M. Bruemmer, PNNL, March 2013. | Download | 3/1/2013 |
| Updates of High-Fluence Induced Microstructural Evolution of Austenitic Stainless Steels Under LWR Relevant Conditions, L. Tan, J. T. Busby, ORNL/TM-2013/84, February 2013 | Updates of High-Fluence Induced Microstructural Evolution of Austenitic Stainless Steels Under LWR Relevant Conditions, L. Tan, J. T. Busby, ORNL/TM-2013/84, February 2013. | Download | 2/1/2013 |
| Cast Stainless Steel Aging Research Plan, T. S. Byun and J. T. Busby, ORNL/LTR-2012/440, September 2012 | Cast Stainless Steel Aging Research Plan, T. S. Byun and J. T. Busby, ORNL/LTR-2012/440, September 2012 | Download | 9/1/2012 |
| LWRS Program Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants, K.L. Simmons, H.M. Hashemian, P. Ramuhalli, R. Konnick, D.L. Brenchley, S. Ray, J.B. Coble, PNNL-21731, September 2012 | LWRS Program Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants, K.L. Simmons, H.M. Hashemian, P. Ramuhalli, R. Konnick, D.L. Brenchley, S. Ray, J.B. Coble, PNNL-21731, September 2012. | Download | 9/1/2012 |
| Light Water Reactor Sustainability (LWRS) Program – R&D Roadmap for Non-Destructive Evaluation (NDE) of Fatigue Damage in Piping, S. Bakhtiari, P. Ramuhalli, D.L. Brenchley, ANL/NE-12/43, September 2012. | Light Water Reactor Sustainability (LWRS) Program – R&D Roadmap for Non-Destructive Evaluation (NDE) of Fatigue Damage in Piping, S. Bakhtiari, P. Ramuhalli, D.L. Brenchley, ANL/NE-12/43, September 2012. | Download | 9/1/2012 |
| Low-temperature Swelling in LWR Internal Components: Current Data and Modeling Assessment, ORNL/LTR-2012/390, R. E. Stoller, A. V. Barashev, and S. I. Golubov, September 2012. | Low-temperature Swelling in LWR Internal Components: Current Data and Modeling Assessment, ORNL/LTR-2012/390, R. E. Stoller, A. V. Barashev, and S. I. Golubov, September 2012. | Download | 9/1/2012 |
| LWRS NDE Workshops Fatigue Workshop Summary, S. Bakhtiari, August 2, 2012. | LWRS NDE Workshops Fatigue Workshop Summary, S. Bakhtiari, August 2, 2012. | Download | 8/2/2012 |
| A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components, S. Mohanty, S. Majumdar, and K. Natesan, Argonne National Laboratory, June 2012. | A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components, S. Mohanty, S. Majumdar, and K. Natesan, Argonne National Laboratory, June 2012. | Download | 6/1/2012 |
| Report on Assessment of Environmentally-Assisted Fatigue for LWR Extended Service Conditions, S. Majumdar and K. Natesan, Argonne National Laboratory, ANL-LWRS-47, September 2011. | Report on Assessment of Environmentally-Assisted Fatigue for LWR Extended Service Conditions, S. Majumdar and K. Natesan, Argonne National Laboratory, ANL-LWRS-47, September 2011. | Download | 9/1/2011 |
| Assessment of Initial Test Conditions for Experiments to Assess Irradiation-Assisted Stress Corrosion Cracking Mechanisms, J. T. Busby and M. N. Gussev, ORNL, ORNL/TM-2010/346, December 2010. | Assessment of Initial Test Conditions for Experiments to Assess Irradiation-Assisted Stress Corrosion Cracking Mechanisms, J. T. Busby and M. N. Gussev, ORNL, ORNL/TM-2010/346, December 2010. | Download | 12/1/2010 |
| Status Update on the Long-Term Crack Initiation Testing of Alloy 690 and Its Weld Metals in PWR Primary Water | Determining the SCC initiation resistance and the aging behavior of high-Cr Ni-base Alloy 690 and its weld metals is needed to confirm their viability as the replacement materials for Alloy 600/182/82 and to ensure safe and economical power production over extended operation of light water reactors. This report provides a progress summary of research activities performed from October 2023 to June 2024 at Pacific Northwest National Laboratory to address these topics. | Download | 07/01/2024 |
| Microstructure and In-Service Degradation of Baffle-Former Bolts – In-Core Components of Light-Water Reactors, ORNL/TM-2023/3118 | Specimen surfaces that are exposed to high-temperature, high-pressure water exhibited signs of in-service corrosion degradation. EBSD and EDS analyses highlighted intergranular corrosion, possible grain boundary oxidation at depths of less than 3 μm, and unexpectedly, short cracks filled with Cr-rich oxides measuring approximately 5–6 μm. | Download | 09/29/2023 |
| Environmentally Assisted Fatigue in Light Water Reactor Environment | Starting from a rich background in theoretical and experimental EAF, ANL previously developed an approach to evaluate fatigue performance of reactor materials in light water reactor environments with the correction factor Fen. The approach was based on a large body of experimental work performed at ANL and elsewhere, and was consistent with American Society of Mechanical Engineers (ASME)’s methodology governing the design and construction of reactor components. In recent years, the program was focused on component fatigue prediction and made several major and fundamental contributions in this area. | Download | 8/14/2024 |
| Report Title | Brief Narrative | Link | Date |
|---|---|---|---|
| Progress on Grizzly Development for Reactor Pressure Vessels and Reinforced Concrete Structures. INL/EXT-19-56012 | The Grizzly code has been under development to provide predictive tools for the evolution of material degradation in critical light water reactor structural components due to long-term exposure to the environmental conditions of normal reactor operation and for the effects of this degradation on the safety of these components. This development has primarily targeted reactor pressure vessels and reinforced concrete structures. | Download | 9/18/2019 |
| Update of modeling effort directed at mitigation approaches to reactor pressure vessel embitterment | Update of modeling effort directed at mitigation approaches to reactor pressure vessel embitterment, S. Cui, M. Mamivand, G. Odette, January 31, 2019. | Download | 1/31/2019 |
| Fracture Toughness Characterization of Highly Irradiated Reactor Pressure Weld from the ATR-2 Experiment, ORNL/TM-2018/918 | The UCSB ATR-2 irradiation experiment aims to generate a comprehensive database on various irradiated RPV steels, addressing a critical knowledge gap in predicting embrittlement at high fluence for extended plant operation of up to 80 years. A primary focus of this experiment is to analyze the effects of irradiation temperature, neutron flux, fluence, and alloy chemistry on the evolution of MNSP and to develop models that illustrate how these factors influence hardening and embrittlement. To facilitate accurate comparisons, the ATR-2 experiment includes DCT specimens of three representative materials, allowing direct measurement of the shift in fracture toughness during the ductile-to-brittle transition. This benchmark data will serve as a reference for evaluating the accuracy of predictive models regarding fracture toughness shifts. | Download | 8/18/2018 |
| The Fracture Toughness Evaluation of Mini-CT specimen Test Results of the Irradiated Midland RPV Beltline Material, ORNL/TM-2018/509 | The Fracture Toughness Evaluation of Mini-CT specimen Test Results of the Irradiated Midland RPV Beltline Material, ORNL/TM-2018/509, M. Sokolov, May 2018. | Download | 5/27/2018 |
| Complete report that documents the completion of a validated model for transition temperature shift in RPV steels September 22, 2017 | Complete report that documents the completion of a validated model for transition temperature shift in RPV steels September 22, 2017 Milestone, S. Shu, M. Mamivand, H. Ke, T. Mayeshiba, B. Afflerbach, J. Ke, D. Morgan, G. Odette, P. Wells, N. Almirall, September 22, 2017. | Download | 9/22/2017 |
| Microstructural Characterization of Reactor Pressure Vessel Alloys: ATR-2 Experiment Update, UCSB ATR-2 2017-2 | Microstructural Characterization of Reactor Pressure Vessel Alloys: ATR-2 Experiment Update, UCSB ATR-2 2017-2, G. Odette, T. Yamamoto, P. Wells, N. Almirall, D. Gragg, September 2017. | Download | 9/20/2017 |
| Microstructural Charaterization of Reactor Pressure Vessel Alloys: ATR-2 Experiment Update, UCSB ATR-2 2017-2 | Microstructural Characterization of Reactor Pressure Vessel Alloys: ATR-2 Experiment Update, UCSB ATR-2 2017-2, G. Odette, T. Yamamoto, P. Wells, N. Almirall, D. Gragg, September 2017. | Download | 9/18/2017 |
| Fracture Toughness Characterization of Reactor Pressure Alloys from the ATR-2 Experiment, ORNL/TM-2017/358 | Fracture Toughness Characterization of Reactor Pressure Alloys from the ATR-2 Experiment, ORNL/TM-2017/358, M. Sokolov, R. Nanstad, G. Odette, T. Yamamoto, P. Wells, July 2017. | Download | 7/18/2017 |
| Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels, ORNL/TM- 2017/275 | Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels, ORNL/TM- 2017/275, M. Sokolov, May 2017. | Download | 5/28/2017 |
| Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials, ORNL/TM-2017/172 | Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials, R. Nanstad, G. Odette, N. Almirall, J. Robertson, L. Server, T. Yamamoto, P. Wells, ORNL/TM-2017/172, April 2017. | Download | 4/1/2017 |
| Complete Report on the Modeling of Precipitate Processes in irradiated Reactor Pressure Vessel Steel | Complete Report on the Modeling of Precipitate Processes in irradiated Reactor Pressure Vessel Steel, M. Mamivand, H. Ke, S. Shu, T. Mayeshiba, D. Morgan, G. Odewtte, P. Wells, N. Almirall, March 31, 2017. | Download | 3/31/2017 |
| Summary of Progress on the ATR-2 Experiment Post-Irradiation Examination of Reactor Pressure Vessel Alloys | Summary of Progress on the ATR-2 Experiment Post-Irradiation Examination of Reactor Pressure Vessel Alloys, UCSB ATR-2 2017-1, G. Odette, T. Yamamoto, P. Wells, N. Almirall, K. Fields, D. Gragg, R. Nanstad, J. Robertson, K. Wilford, N. Riddle and T. Williams, March 31, 2017. | Download | 3/31/2017 |
| The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for Fracture Toughness Master Curves for Reactor Pressure Vessel Steels, ORNL/TM-2016/602 | The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for Fracture Toughness Master Curves for Reactor Pressure Vessel Steels, ORNL/TM-2016/602, M. Sokolov, R. Nanstad, September 2016. | Download | 9/1/2016 |
| Zion Unit 1 Reactor Pressure Vessel Sample Acquisition: Phase 2 and Phase 3 Status Report , ORNL/TM-2016/561 | Zion Unit 1 Reactor Pressure Vessel Sample Acquisition: Phase 2 and Phase 3 Status Report, ORNL/TM-2016/561, T. Rosseel, M. Sokolov, X. Chen, R. Nanstad, September 2016. | Download | 9/1/2016 |
| Update on the High Fluence Advanced Test Reactor – 2 Reactor Pressure Vessel High Fluence Irradiation Project | Update on the High Fluence Advanced Test Reactor – 2 Reactor Pressure Vessel High Fluence Irradiation Project, UCSB ATR-2 2016-1, G. Odette, T. Yamamoto, P Wells, N. Almirall, K. Fields, D. Gragg, R. Nanstad, M. Sokolov, J. Robertson, June 30, 2016 | Download | 6/30/2016 |
| Report on the Harvesting and Acquisition of Zion Unit 1 Reactor Pressure Vessel Segments, ORNL/TM-2016/240 | Report on the Harvesting and Acquisition of Zion Unit 1 Reactor Pressure Vessel Segments, ORNL/TM-2016/240, T. Rosseel, M. Sokolov, R. Nanstad, June 2016. | Download | 6/1/2016 |
| Modeling Late Blooming Phase Evolution during Post-irradiation Annealing in Select Reactor Pressure Vessels: Milestone 2 Report | Modeling Late Blooming Phase Evolution during Post-irradiation Annealing in Select Reactor Pressure Vessels: Milestone 2 Report, S. Shu, H. Ke, D. Morgan, G. Odette, P. Wells, March 2016. | Download | 3/31/2016 |
| Results of Fracture Toughness Tests for the Round Robin Test Program Using Mini-Compact Specimens, ORNL/LTR-2014/686 | Results of Fracture Toughness Tests for the Round Robin Test Program Using Mini-Compact Specimens, M.A. Sokolov, Oak Ridge National Laboratory, ORNL/LTR-2014/686, December 2014. | Download | 12/1/2014 |
| Status Report Describing Evaluation of Embrittlement Effects in a Reactor Pressure Vessel, | Status Report Describing Evaluation of Embrittlement Effects in a Reactor Pressure Vessel Nozzle, M. Backman, B. Spencer, R. Dodds, B. Wirth, and J. Busby, August 2014. | Download | 8/1/2014 |
| Comprehensive and Comparative Analysis of Atom Probe, Small-Angle Neutron Scattering, and Other Microstructural Experiments on Available High Fluence Reactor Pressure Vessel Steels | Comprehensive and Comparative Analysis of Atom Probe, Small-Angle Neutron Scattering, and Other Microstructural Experiments on Available High Fluence Reactor Pressure Vessel Steels, M. Sokolov, R. Nanstad, M. Miller, K. Littrell, G. Odette, P. Wells, D. Sprouster, and L. Ecker, ORNL/TM-2014/241, June 2014. | Download | 6/1/2014 |
| Grizzly/FAVOR Interface Project Report, ORNL/TM-2013/44094, | Grizzly/FAVOR Interface Project Report, T.L. Dickson, P.T. Williams, S. Yin, H.B. Klasky, S. K. Tadinada, B.R. Bass, ORNL/TM-2013/44094, June 2013. | Download | 6/1/2013 |
| Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone M2LW-13OR0402013, Progress Report on Status of Advanced Test Reactor-2 Reactor Pressure Vessel Materials Irradiation Project, R. K. Nanstad, G. R. Odette, ORNL/LTR-2013/ | Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone M2LW-13OR0402013, Progress Report on Status of Advanced Test Reactor-2 Reactor Pressure Vessel Materials Irradiation Project, R. K. Nanstad, G. R. Odette, ORNL/LTR-2013/129, March 2013. | Download | 3/1/2013 |
| Report on Small-Angle Neutron Scattering Experiments of Irradiated Reactor Pressure Vessel Materials, M. A. Sokolov, K. C. Littrell, R. K. Nanstad, ORNL/TM-2012/630, December 2012 | Report on Small-Angle Neutron Scattering Experiments of Irradiated Reactor Pressure Vessel Materials, M. A. Sokolov, K. C. Littrell, R. K. Nanstad, ORNL/TM-2012/630, December 2012. | Download | 12/1/2012 |
| Roadmap for Nondestructive Evaluation (NDE) of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program, C. Smith, R. Nanstad, R. Odette, D. Clayton, K. Matlack, P. Ramuhalli. G. Light, ORNL/TM-2012/380, September | Roadmap for Nondestructive Evaluation (NDE) of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program, C. Smith, R. Nanstad, R. Odette, D. Clayton, K. Matlack, P. Ramuhalli. G. Light, ORNL/TM-2012/380, September 2012. | Download | 9/1/2012 |
| LWRS Non-Destructive Evaluation (NDE) workshop Summary for Reactor Pressure Vessels (RPV), C.W. Smtih, August 9, 2012 | LWRS Non-Destructive Evaluation (NDE) workshop Summary for Reactor Pressure Vessels (RPV), C.W. Smith, August 9, 2012. | Download | 8/9/2012 |
| Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Letter Report on Metallurgical Examination of the High Fluence RPV Specimens From the Ringhals Nuclear Reactors, R.K. Nanstad, Oak Ridge National Laboratory, ORNL/LTR-2012/113, Ma | Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Letter Report on Metallurgical Examination of the High Fluence RPV Specimens From the Ringhals Nuclear Reactors, R.K. Nanstad, Oak Ridge National Laboratory, ORNL/LTR-2012/113, March 2012. | Download | 3/1/2012 |
| Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges, R. K. Nanstad, ORNL, and W. L. Server, ATI Consulting, ORNL/LTR-2011/351, September 2011. | Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges, R. K. Nanstad, ORNL, and W. L. Server, ATI Consulting, ORNL/LTR-2011/351, September 2011. | Download | 9/1/2011 |
| Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials, R. K. Nanstad, ORNL, ORNL/LTR-2011/172, June 2011. | Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials, R. K. Nanstad, ORNL, ORNL/LTR-2011/172, June 2011. | Download | 6/1/2011 |
| Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment, R. K. Nanstad, ORNL, and G. R. Odette, University of California, Santa Barbara, ORNL/LTR-2011/41 | Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment, R. K. Nanstad, ORNL, and G. R. Odette, University of California, Santa Barbara, ORNL/LTR-2011/413, January 2011. | Download | 1/1/2011 |
