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Nondestructive Evaluation (NDE) of Cable Moisture Exposure using Frequency Domain Reflectometry (FDR)124035Nondestructive Evaluation (NDE) of Cable Moisture Exposure using Frequency Domain Reflectometry (FDR), PNNL-31934, S. Glass, M. Spencer, A. Sriraman, L. Fifield, M. Prowant, September 2021.9/17/2021 7:54:19 AMLight Water Reactor Sustainability Program Nondestructive Evaluation (NDE) of Cable Moisture Exposure using Frequency Domain Reflectometry (FDR) September 2021 U.S. Department 27https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
Sequential Versus Simultaneous Aging of XLPE and EPDM Nuclear Cable Insulation Subjected to Elevated Temperature and Gamma Radiation (Final Results)97503Sequential Versus Simultaneous Aging of XLPE and EPDM Nuclear Cable Insulation Subjected to Elevated Temperature and Gamma Radiation (Final Results), PNNL-30041 Rev. 1, L. Fifield, M. Spencer, Y. Ni, D. Li, M. Pallaka, T. Bisel, A. Zwoster, M. Murphy, December 2020.12/11/2020 10:34:11 PMU.S. Department of Energy Office of Nuclear Energy This information was prepared as an account of work sponsored by an agency of the U.S. Government This report addresses one of the 148https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
Evaluation of Critical Parameters to Model Stress Corrosion Crack Initiation in Alloy 600 and Alloy 182 in PWR Primary Water200485Evaluation of Critical Parameters to Model Stress Corrosion Crack Initiation in Alloy 600 and Alloy 182 in PWR Primary Water, M2LW-20OR0402036, Z. Zhai, J. Wang, M. Toloczko, S. Bruemmer, September 2020.9/15/2020 6:13:28 PMU.S. Department of Energy Office of Nuclear Energy This information was prepared as an account of work sponsored by an agency of the U.S. Government In this report, a brief summary 139https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
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, 200486Evaluation 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, G. Was, D. Du, September 2020. 9/28/2020 7:16:42 PMM2LW-20OR0402023 Light Water Reactor Sustainability Program Evaluation of the stress and fluence dependence of irradiation assisted stress corrosion crack initiation in high 123https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
Potential Life Extension Strategies for In-Service Degraded Cables200487Potential Life Extension Strategies for In-Service Degraded Cables, PNNL-30402, S. Nune, M. Spencer, L. Fifield, September 2020.9/17/2020 12:42:36 AMU.S. Department of Energy Office of Nuclear Energy This information was prepared as an account of work sponsored by an agency of the U.S. Government As cable materials age over time 166https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
Intermediate-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at the LWR Relevant Temperature200488Intermediate-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at the LWR Relevant Temperature, ORNL/TM-2020/1754, L. Tan, X. Chen, September 30, 2020.9/27/2020 3:05:21 PMReports produced after January 1, 1996, are generally available free via US Department of Energy (DOE) SciTech Connect Reports produced before January 1, 1996, may be purchased by 169https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
A Hybrid AI/ML and Computational Mechanics Based Approach for Time-Series State and Fatigue Life Estimation of Nuclear Reactor Components200489A Hybrid AI/ML and Computational Mechanics Based Approach for Time-Series State and Fatigue Life Estimation of Nuclear Reactor Components, ANL/LWRS-20/01, S. Mohanty, J. Listwan, September 2020.9/22/2020 12:17:05 AMArgonne is a U.S. Department of Energy laboratory managed by UChicago Argonne, LLC under contract DE-AC02-06CH11357 The Laboratory’s main facility is outside Chicago, at 9700 149https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
Analysis of Deformation Localization Mechanisms in Highly Irradiated Austenitic Stainless Steel via In Situ Techniques200490Analysis of Deformation Localization Mechanisms in Highly Irradiated Austenitic Stainless Steel via In Situ Techniques,ORNL/TM-2020/1739, M. Gussev, N. Bibhanshu, E. Cakmak, J. Dixon, September 2020.9/24/2020 3:01:38 AMReports produced after January 1, 1996, are generally available free via US Department of Energy (DOE) SciTech Connect Reports produced before January 1, 1996, may be purchased by 194https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
Quantifying micro-galvanic corrosion in stainless steels activated by post-yielding microstructures200491Quantifying micro-galvanic corrosion in stainless steels activated by post-yielding microstructures, M2LW-20R0402027, X. Chen, M. Gussev, G. Sant, September 2020.9/15/2020 5:11:01 PMU.S. Department of Energy Office of Nuclear Energy This information was prepared as an account of work sponsored by an agency of the U.S. Government This research was carried out in 110https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf
Assessment of Grizzly Capabilities for Reactor Pressure Vessels and Reinforced Concrete Structures200492Assessment of Grizzly Capabilities for Reactor Pressure Vessels and Reinforced Concrete Structures, INL/EXT-20-00617, B. Spencer, W. Hoffman, S. Biswas, A. Jain, September 2020.9/15/2020 3:24:29 PMINL/EXT-20-00617 Light Water Reactor Sustainability Program Assessment of Grizzly Capabilities for Reactor Pressure Vessels and Reinforced Concrete Structures Benjamin W 123https://lwrs.inl.gov/Materials Aging and Degradation/Forms/AllItems.aspxpdfFalsepdf