| 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. | 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 | 4/2/2024 | | Concrete Aging and Degradation | Concrete | Materials Research | |
| 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. | Assessment of Machine Learning for Ultrasonic Nondestructive Evaluation of Alkali–Silica Reaction in Concrete, ORNL/TM-2024/3295 | 3/11/2024 | | Concrete Aging and Degradation | Concrete | Materials Research | |
| 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. | Microstructure and In-Service Degradation of Baffle-Former Bolts – In-Core Components of Light-Water Reactors, ORNL/TM-2023/3118 | 9/29/2023 | | Reactor Core and Primary Systems | Reactor Core and Primary Systems | Materials Research | |
| 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 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. | Spread Spectrum Time Domain Reflectometry (SSTDR) and Frequency Domain Reflectometry (FDR) for Detection of Cable Anomalies Using Machine Learning, PNNL-34821 | 9/26/2023 | | Reactor Core and Primary Systems | Cable Aging and Cable NDE | Materials Research | |
| 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. | Extended Bandwidth Spread Spectrum Time Domain Reflectometry Cable Test for Thermal Aging, Low Resistance Fault, and Water Detection, PNNL-34815 | 9/26/2023 | | Reactor Core and Primary Systems | NDE of Cable and Cable Insulation | Materials Research | |
| 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. | The Mechanism of Irradiation Assisted Stress Corrosion Cracks in Stainless Steels, M2LW-23OR0402029 | 9/26/2023 | | Reactor Core and Primary Systems | Reactor Metals | Materials Research | |
| 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. | Complete the additional microstructural evaluation and SCC CGR testing on two heats of aged Alloy 152, ANL/LWRS-23/1 | 9/26/2023 | | Reactor Core and Primary Systems | Reactor Metals | Materials Research | |
| In nuclear power plants, 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. | Applying grain-boundary sensitive electrochemical scanning probe techniques to evaluate intergranular degradation of irradiated and deformed stainless steels, AM3LW-23OR04020211 | 9/26/2023 | | Reactor Core and Primary Systems | Reactor Core and Primary Systems | Materials Research | |
| 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. | 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 | 9/15/2023 | | Mitigation Methods and Technologies | Mitigation Methods and Technologies | Materials Research | |
| 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. | Status of Cable Aging Knowledge Gaps Identified in the Expanded Materials Degradation Assessment (EMDA) | 9/15/2023 | | Cable Aging and Cable NDE | Cables | Materials Research | |
| 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. | Comprehensive Characterization of Helium-induced Degradation of the Friction Stir Weld on Neutron Irradiated 304L Stainless Steel, ORNL/TM-2023/3088 | 9/12/2023 | | Mitigation Methods and Technologies | Mitigation Methods and Technologies | Materials Research | |
| 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. The material types and the specific water chemistry conditions evaluated in this research were selected based on discussions with EPRI, who is assisting the U.S. pressurized water reactor utilities in a potential transition from LiOH to KOH. | FY 2023 Progress on Stress Corrosion Crack Testing of Ni-Base Alloys in PWR Primary Water, M2LW-23OR04020311 | 8/29/2023 | | Reactor Core and Primary Systems | Reactor Core and Primary Systems | Materials Research | |
| 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. | Survey of Aging and Monitoring Concerns for Cables and Splices Due to Cable Repair and Replacement, PNNL-34576 | 7/31/2023 | | Cable Aging and Cable NDE | Cables | Materials Research | |
| 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. | Machine Learning for Processing Ultrasonic Data from Long-Term Monitoring of Concrete with Alkali-Silica Reaction (ASR) | 6/29/2023 | | Concrete Aging and Degradation | Concrete | Materials Research | |
| Operating experience of Type 304(L)/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. | Preparation for Stress Corrosion Crack Initiation Testing of Austenitic Stainless Steels in PWR Primary Water | 6/26/2023 | | Reactor Core and Primary Systems | Reactor Core and Primary Systems | Materials Research | |
| 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. | Frequency Domain Reflectometry (FDR) Simulation Techniques for Digital Twin Representation of an Electrical Cable, Milestone M3LW-23OR0404023 | 4/14/2023 | | Cable Aging and Cable NDE | Cables | Materials Research | |
| 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. | Dose Rate Effects on Degradation of Nuclear Power Plant Electrical Cable Insulation at a Common Dose. PNNL-34068 | 3/24/2023 | | Cable Aging and Cable NDE | Reactor Core and Primary Systems | Materials Research | |
| 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. | Microstructural characterizations of the second high fluence baffle-former bolt retrieved from a Westinghouse two-loop downflow type PWR, ORNL/TM-2022/2668 | 10/10/2022 | | Reactor Core and Primary Systems | Reactor Metals | Materials Research | |
| The results of the first laser welding campaign on irradiated Ni-base alloy 182 and the preliminary weld quality inspections at the Radiochemical Engineering Development Center (REDC). The irradiated nickel alloy 182, with target helium contents of 5 appm, 10 appm, and 20 appm, were successfully laser welded in the hot cell. The significant, on-going effort to weld irradiated alloys with high helium concentrations and comprehensively analyze the results will yield validated repair techniques and guidelines for use by the nuclear industry in extending the operational lifetimes of nuclear power plants. | Complete the weld campaign on Ni-base irradiated materials using stress improved laser welding including the preliminary weld quality inspections, ORNL/SPR-2022/2657 | 9/28/2022 | | Mitigation Methods and Technologies | Mitigation Methods and Technologies | Materials Research | |
| 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. | 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 | 9/26/2022 | | Reactor Core and Primary Systems | Mechanisms of IASCC | Materials Research | |
| 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. | Electrochemical probing of microstructural heterogeneities in irradiated and deformed stainless steel, M3LW-22OR0402028 | 9/18/2022 | | Reactor Core and Primary Systems | Mechanisms of IASCC | Materials Research | |
| 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. | 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 | 9/17/2022 | | Reactor Core and Primary Systems | Reactor Core and Primary Systems | Materials Research | |
| 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. | Effect of thermal aging on microstructure and crack growth response of Alloy 152 Weld, ANL/LWRS-22/2 | 9/10/2022 | | Reactor Core and Primary Systems | Reactor Core and Primary Systems | Materials Research | |
| 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. | Inverse Temperature Effects in Nuclear Power Plant Electrical Cable Insulation, PNNL-33296 | 9/9/2022 | | Cable Aging and Cable NDE | Cables | Materials Research | |
| 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. | Complete the stress corrosion crack initiation and crack growth response of Ni-base alloys in KOH vs. LiOH PWR primary water chemistry, M3LW-22OR0402033 | 7/29/2022 | | Reactor Core and Primary Systems | Concrete | Materials Research | |
| The U.S. nuclear industry is considering replacing LiOH with potassium hydroxide (KOH) for pH control in 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 SCC response of Ni-base alloys in a KOH environment to ensure that SCC susceptibility is not increased by KOH water chemistry. | Stress Corrosion Cracking of Ni-base Alloys in PWR Primary Water Containing KOH vs. LiOH, M3LW-22OR0402033 | 7/11/2022 | | Reactor Core and Primary Systems | Concrete | Materials Research | |
| 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. | Reconstruction of 3D Concrete Microstructures Combining High-Resolution Characterization and Convolutional Neural Network for Image Segmentation | 6/12/2022 | | Concrete Aging and Degradation | Concrete Aging and Degradation | Materials Research | |
| 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. | Ultrasonic Model Based Iterative Reconstruction of Experimental Concrete Specimens at EPR, ORNL/SPR-2022/2318 | 1/1/2022 | | Concrete Aging and Degradation | Concrete | Materials Research | |
| 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. | 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 | 9/27/2021 | | Reactor Core and Primary Systems | | Materials Research | |
| 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. | 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 | 9/22/2021 | | Reactor Core and Primary Systems | | Materials Research | |