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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Fleet-Wide Prognostic and Health Management Suite: Asset Fault Signature Database

Proactive online monitoring in the nuclear industry is being explored using the Electric Power Research Institute’s Fleet-Wide Prognostic and Health Management (FW-PHM) Suite software. The FW-PHM Suite is a set of web-based diagnostic and prognostic tools and databases that serves as an integrated health monitoring architecture. The FW-PHM Suite has four main modules: (1) Diagnostic Advisor, (2) Asset Fault Signature (AFS) Database, (3) Remaining Useful Life Advisor, and (4) Remaining Useful Life Database. The paper focuses on the AFS Database of the FW-PHM Suite, which is used to catalog asset fault signatures. A fault signature is a structured representation of the information that an expert would use to first detect and then verify the occurrence of a specific type of fault. The fault signatures developed to assess the health status of generator step-up transformers are described in the paper. The developed fault signatures capture this knowledge and implement it in a standardized approach, thereby streamlining the diagnostic and prognostic process. This will support the automation of proactive online monitoring techniques in nuclear power plants to diagnose incipient faults, perform proactive maintenance, and estimate the remaining useful life of assets.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Poster — Thur Eve — 09: Evaluation of electrical impedance and computed tomography fusion algorithms using an anthropomorphic phantom

Integration of biological conductivity information provided by Electrical Impedance Tomography (EIT) with anatomical information provided by Computed Tomography (CT) imaging could improve the ability to characterize tissues in clinical applications. In this paper, we report results of our study which compared the fusion of EIT with CT using three different image fusion algorithms, namely: weighted averaging, wavelet fusion, and ROI indexing. The ROI indexing method of fusion involves segmenting the regions of interest from the CT image and replacing the pixels with the pixels of the EIT image. The three algorithms were applied to a CT and EIT image of an anthropomorphic phantom, constructed out of five acrylic contrast targets with varying diameter embedded in a base of gelatin bolus. The imaging performance was assessed using Detectability and Structural Similarity Index Measure (SSIM). Wavelet fusion and ROI-indexing resulted in lower Detectability (by 35% and 47%, respectively) yet higher SSIM (by 66% and 73%, respectively) than weighted averaging. Our results suggest that wavelet fusion and ROI-indexing yielded more consistent and optimal fusion performance than weighted averaging.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

Systematic Enterprise Risk Management by Integrating the RISMC Toolkit and Cost-Benefit Analysis (Final Report)

The goal of this research is to theorize and quantify the relationships between safety and the financial performance of nuclear power plants (NPPs). The Socio-Technical Risk Analysis (SoTeRiA) theoretical framework, which connects the social aspects (e.g., safety culture) and structural features (e.g., safety practices) of an organization with organizational safety and financial risks, is used to theorize the direct and indirect relationships between safety and the financial performance of NPPs. An Integrated Enterprise Risk Management (I-ERM) methodological framework is developed to operationalize SoTeRiA to quantify NPP safety and financial performance in a unified platform where their underlying physical degradation mechanisms, coupled with maintenance performance (considering human and organizational factors), are explicitly incorporated to depict the interconnections and dependencies between safety and financial performance. In this study, NPP safety refers to both occupational safety and system safety (estimated from Probabilistic Risk Assessment, PRA), and financial performance refers to the monetary values associated with NPP operation and maintenance (O&M) strategies. This report covers a case study demonstrating the feasibility of the I-ERM methodological framework. More detailed development of one of the I-ERM modules, i.e., Probabilistic Physics-of-Failure (PPoF) analysis, and its connection with other I-ERM modules is demonstrated in a second case study. The outcome of this research will help NPP decision-makers create cost-saving maintenance strategies while maintaining safety by providing cost- and risk-informed recommendations regarding maintenance work processes and operational strategies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Big Data Framework with Machine Learning for D and D Applications - 19108

The nuclear industry is experiencing a steady increase in maintenance costs even though plants are maintained under high levels of safety, capability and reliability. Nuclear power plants are expected to run every unit at maximum capacity at all times, efficiently utilizing assets with minimal downtime. Surveillance and maintenance of nuclear decommissioning infrastructure provides lot challenges with respect to maintenance or decommissioning of the buildings. There is a need for a framework to analyze the huge amount of data generated by the sensors on the nuclear reactor components as well as structures, to monitor the conditions of these building over a period of time. Emerging technologies such as big data analytics have become a requirement in the nuclear industry to improve structural health monitoring and diagnostics. FIU will make use of existing mature technologies in the areas of imaging, robotics, big data, and machine learning/deep learning to assess the structural integrity of aging facilities at DOE sites. As these facilities await decommissioning, there is a need to understand the structural health of these structures. Many of these facilities were built over 50 years ago and in some cases these facilities have gone beyond the operational life expectancy. In other cases, the facilities have been placed in a state of 'cold and dark' and they are sitting unused, awaiting decommissioning. Finally, some older facilities are one-of-a-kind operational/production facilities supporting critical DOE missions and cannot be shut down. In any of these scenarios, the structural integrity of these facilities may be compromised, so it is imperative that adequate inspections and data collection/analysis be performed on a continuous and ongoing basis. The primary goals of the research include collecting various formats of data such as structured, unstructured and streaming data from the various sensors deployed in buildings, collect videos/pictures from various imaging sources, ingest them using a Hadoop distributed file system and process the data using Spark to perform batch processing and real time analytics. Research and development on various machine learning/deep learning algorithms will also be performed to analyze the heterogeneous data collected from nuclear decommission infrastructure. FIU will design the big data framework to ingest, store and process huge amounts of heterogeneous data collected from many sources and optimize the algorithms to provide insights into the data and predict anomalies observed when compared against baseline conditions. Various modules of this framework will include heterogeneous data sources, message broker, Hadoop distributed file system, Spark for stream and batch processing, machine learning/ deep learning, Cassandra - persistent data store and visualization. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Improving Site-Specific Radiological Performance Assessments - 13431

An improved approach is presented for conducting complete and defensible radiological site-specific performance assessments (PAs) to support radioactive waste disposal decisions. The basic tenets of PA were initiated some thirty years ago, focusing on geologic disposals and evaluating compliance with regulations. Some of these regulations were inherently probabilistic (i.e., addressing uncertainty in a quantitative fashion), such as the containment requirements of the U.S. Environmental Protection Agency's (EPA's) 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, Chap. 191.13 [1]. Methods of analysis were developed to meet those requirements, but at their core early PAs used 'conservative' parameter values and modeling approaches. This limited the utility of such PAs to compliance evaluation, and did little to inform decisions about optimizing disposal, closure and long-term monitoring and maintenance, or, in general, maintaining doses 'as low as reasonably achievable' (ALARA). This basic approach to PA development in the United States was employed essentially unchanged through the end of the 20. century, principally by the U.S. Department of Energy (DOE). Performance assessments developed in support of private radioactive waste disposal operations, regulated by the U.S. Nuclear Regulatory Commission (NRC) and its agreement states, were typically not as sophisticated. Discussion of new approaches to PA is timely, since at the time of this writing, the DOE is in the midst of revising its Order 435.1, Radioactive Waste Management [2], and the NRC is revising 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste [3]. Over the previous decade, theoretical developments and improved computational technology have provided the foundation for integrating decision analysis (DA) concepts and objective-focused thinking, plus a Bayesian approach to probabilistic modeling and risk analysis, to guide improvements in PA. This decision-making approach, [4, 5, 6] provides a transparent formal framework for using a value- or objective-focused approach to decision-making. DA, as an analytical means to implement structured decision making, provides a context for both understanding how uncertainty affects decisions and for targeting uncertainty reduction. The proposed DA approach improves defensibility and transparency of decision-making. The DA approach is fully consistent with the need to perform realistic modeling (rather than conservative modeling), including evaluation of site-specific factors. Instead of using generic stylized scenarios for radionuclide fate and transport and for human exposures to radionuclides, site-specific scenarios better represent the advantages and disadvantages of alternative disposal sites or engineered designs, thus clarifying their differences as well as providing a sound basis for evaluation of site performance. The full DA approach to PA is described, from explicitly incorporating societal values through stakeholder involvement to model building. Model building involves scoping by considering features, events, processes, and exposure scenarios (FEPSs), development of a conceptual site model (CSM), translation into numerical models and subsequent computation, and model evaluation. These are implemented in a cycle of uncertainty analysis, sensitivity analysis and value of information analysis so that uncertainty can be reduced until sufficient confidence is gained in the decisions to be made. This includes the traditional focus on hydrogeological processes, but also places emphasis on other FEPSs such as biotically-induced transport and human exposure phenomena. The significance of human exposure scenarios is emphasized by modifying the traditional acronym 'FEPs' to include them, hence 'FEPSs'. The radioactive waste community is also recognizing that disposal sites are to be considered a national (or even global) resource. As such, there is a pressing need to optimize their utility within the constraints of protecting human health and the environment. Failing to do so will result in the need for additional sites or options for storing radioactive waste temporarily, assuming a continued need for radioactive waste disposal. Optimization should be performed using DA, including economic analysis, invoked if necessary through the ALARA process. The economic analysis must recognize the cost of implementation (disposal design, closure, maintenance, etc.), and intra- and inter-generational equity in order to ensure that the best possible radioactive waste management decisions are made for the protection of both current and future generations. In most cases this requires consideration of population or collective risk. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Assessment of Measurement Uncertainties in the Jupiter High-240 Experiment

The Jupiter High-240 experiment performed in May of 2019 was previously discussed as a variant of the original Jupiter experiment incorporating plutonium metal alloy fuel plates with higher 240Pu content and lead plates, using both a reference configuration and a second configuration where eight lead plates were replaced with aluminum to simulate voiding. Measurements were recorded for experiment period, the “pressure” of the Comet ram upon closure for each near-critical measurement, and temperature. The experiment reactor period is the time it would take to increase the neutron population by a factor of e. For this experiment, the copper reflectors and upper third of the fuel sits upon a support structure with the lower fuel arrays raised up into the center of the reflectors using a ram (see Fig. 1). The recorded logbook temperature for each measurement corresponds to a resistance temperature detector (RTD) located at the top center of the upper fuel array. This paper summarizes the evaluated uncertainties for the Jupiter High 240 experiment as contributed via the recorded measurements and nuclear data and their assessed impact upon the computation of system reactivity and eigenvalue.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Feasibility Assessment of Using Electrical Impedance Tomography for Damage Localization in Graphite Microreactor Components

Microreactors have great potential to decrease capital costs and construction timelines, reducing the barriers to implementing advanced nuclear reactor technologies. However, the lower power output of these microreactors introduces economic challenges if the operation and maintenance costs cannot be reduced sufficiently. Many microreactor concepts use graphite materials for in-core moderator and structural purposes, which will require periodic inspection or, ideally, in situ structural health monitoring. This work describes an initial evaluation of the feasibility of leveraging graphite’s semiconducting properties to perform electrical impedance tomography (EIT) for defect localization. First, a study was performed to identify the best methods for bonding electrical contacts to minimize the effects of contact resistance that interfere with impedance measurements of the graphite. After determining the best bonding approach, electrodes were attached to multiple graphite components with varying geometries (e.g., simple block and more representative microreactor hexagonal block). In parallel, finite element analysis approaches were developed and implemented to analyze the impact of defects and contact resistances and to inform an iterative inverse model for reconstructing the conductivity distribution. The results of this study show that defect localization in graphite components is possible using EIT if a sufficient number of electrodes (to improve spatial resolution) can be bonded using a technique with very low contact resistance and those contact resistances remain stable during reactor operation. If the reactor core and vessel design do not allow electrodes to be bonded during reactor operation, then it may be possible to detach/reattach electrodes between reactor operation and outages if a suitable mechanical connection that offers repeatable contact resistance can be identified. Alternatively, future work could focus on identifying contacts that can survive the harsh reactor operating environment so that contacts do not need to be removed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Multiphysics Design Optimization and Additive Manufacturing of Nuclear Components (Final CRADA Report - Executive Summary)

Westinghouse Electric Company (WEC) actively participated in the advancement of the nuclear fuel and reactor design space and requested the help of Oak Ridge National Laboratory (ORNL) in the creation of a new design tool set. This report details the creation of a collection of software tool sets that are linked together to collectively assist WEC design engineers in developing novel ideas outside the normal scope of traditional nuclear fuel and reactor design formulas. Specifically, Siemens HEEDS, a design space exploration and parametric optimization software, monitored and changed parameters in a collection of softwares to meet the team’s objective. The HEEDS parametric optimization method, SHERPA, was developed to control the Siemens NX CAD platform to adjust the native CAD of a hexahedral spacer grid. This new geometry can be used to execute a topological design optimization by the NX Topology software add-in. The resulting geometry is additively manufacturable. This topological optimization occurred twice—once on the spacer grid’s spring, and once on the dimple geometry. These new geometries were imported by Siemens’ STAR-CCM+, a multiphysics structural and fluid dynamic computational solver in which the spring geometry is deflected to match the rod insertion configuration. Along with the dimple geometry, this new deflected spring was used to complete a hydraulic assessment of a single-unit cell comprising one rod, one spring, and two dimples. The HEEDS SHERPA algorithm ranks the design based on the final mass of the unit cell and the hydraulic pressure drop performance. The ORNL team demonstrated the ability to use this software and provided engineering judgement to apply modern aerospace aerodynamic design. The effort has been focused on thinking outside the conventional design space and redesigning a spacer grid to perform beyond the WEC set objectives. Furthermore, the ORNL team also demonstrated that the HEEDS optimization routine can independently develop a design that meets the WEC design goals. Although these designs were at a low technology readiness level, their demonstration confirmed the team’s capability to create novel advanced nuclear concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Waste Chemistry Envelope for the Determination of Pitting Corrosion Susceptibility - 19554

Liquid radioactive wastes at the Hanford and Savannah River sites are stored in large underground carbon steel tanks. The high-level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks, will be required to be in service for times exceeding the initial intended life (i.e., greater than 60 years). Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. Administrative programs are in place to mitigate potential corrosion mechanisms and thereby maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. Waste chemistry and temperature control are the means by which corrosion degradation of the waste tanks is minimized. Chemistry controls are based on in-situ coupon tests, waste tank experience, and laboratory testing. These requirements comprise the technical standards for corrosion control. The standards are utilized then as guidelines for operational procedures. The present waste chemistry controls for the Hanford DSTs were established in the 1980's in response to tank failures caused by stress corrosion cracking (SCC) at the DOE Savannah River Site (SRS). Portions of the SRS chemistry limits, which were developed at the Savannah River National Laboratory (SRNL) were coupled with corrosion testing results from Pacific Northwest National Laboratories (PNNL) and adopted for corrosion control of the double-shell tank (DST) waste at the Hanford Site. Since that time, the temperatures of the waste in the DSTs have decreased to less than 50 deg. C in most cases as the radioactive material has decayed. As a consequence, new chemistry controls were formulated for SCC in 2010. This testing however, did not address that potential for pitting in the supernatant and at the liquid-air interface. Indeed, initial tests indicated that the inhibitor requirements for SCC would be insufficient for the prevention of pitting. In addition to lower waste temperatures, future retrieval and process changes may impact the DST waste chemistry. Initial projections suggest that the waste chemistry of the DSTs may shift to a broader range of pH and higher aggressive anion concentrations (e.g., chloride, sulfate, etc.) than the wastes that are presently stored. Additionally, processes within the vitrification facility may result in streams that are returned to the tank farm for evaporation. These streams may also contain high concentrations of aggressive species. SRNL investigated the propensity for pitting of the carbon steel waste tanks at current and projected DST waste chemistries. A statistically designed series of tests provided a pitting factor that relates the ratio of the inhibitor species to the aggressive species and the probability of observing pitting. An acceptable pitting factor was determined and was utilized for the development of a new chemistry envelope. The new chemistry envelope was compared to the envelope that was established for SCC and was determined to be bounding. Validation of the new chemistry envelope was performed via model simulations and comparison with historical experimental results and approaches to establishing chemistry envelopes. The new chemistry envelope will eventually be part of an overall assessment of the condition of the tanks and any changes made to the overall structural integrity program for the Hanford DSTs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

The conceptual solutions concerning decommissioning and dismantling of Russian civil nuclear powered ships

From 1959 up to 1991 nine civil nuclear powered ships were built in Russia: eight ice-breakers and one lash lighter carrier (cargo ship). At the present time three of them were taking out of service: ice-breaker 'Lenin' is decommissioned as a museum and is set for storage in the port of Murmansk, nuclear ice-breakers 'Arktika' and 'Sibir' are berthing. The ice-breakers carrying rad-wastes appear to be a possible source of radiation contamination of Murmansk region and Kola Bay because the ship long-term storage afloat has the negative effect on hull's structures. As the result of this under the auspices of the Federal Targeted Program 'Nuclear and Radiation Safety of Russia for 2008 and the period until 2015' the conception and projects of decommissioning of nuclear-powered ships are developed by the State corporation Rosatom with the involvement of companies of United Shipbuilding Corporation. In developing the principal provisions of conception of decommissioning and dismantling of icebreakers the technical and economic assessment of dismantling options in ship-repairing enterprises of North-West of Russia was performed. The paper contains description of options, research procedure, analysis of options of decommissioning and dismantling of nuclear ice-breakers, taking into account the principle of optimization of potential radioactive effect to personnel, human population and environment. The report's conclusions contain the recommendations for selection of option for development of nuclear icebreaker decommissioning and dismantling projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Concrete in engineered barriers for radioactive waste disposal facilities - phenomenological study and assessment of long term performance

The paper aims to highlight recent developments at the Belgian Nuclear Research Centre SCK-CEN in experimental and numerical study of the coupled physical-chemical behaviour of concrete subject to chemical degradation. The discussion mainly focusses on three inter-linked research projects covering novel experimental methods to study the alteration of hydraulic and transport properties during carbonation and calcium leaching, a pore scale numerical model to capture microstructural changes due to the above degradation processes and a generic multiscale model aimed at determining evolution of the properties of a macrostructure over the long term. The paper also describes supplementary continuum scale numerical studies concerning concrete-clay interactions and geochemical impact on the physical structure of concrete. Preliminary findings from these studies show encouraging results such as the development of novel leaching, water permeability and diffusion apparatus, a robust pore scale model based on Lattice-Boltzmann method and a mesoscale study focused on the importance of interfacial transition zones on the effective diffusivity for linear and nonlinear diffusion problems. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Light Water Reactor Sustainability Non-Destrustive Evaluation for Concrete Research and Development Roadmap

Materials issues are a key concern for the existing nuclear reactor fleet as material degradation can lead to increased maintenance, increased downtown, and increased risk. Extending reactor life to 60 years and beyond will likely increase susceptibility and severity of known forms of degradation. Additionally, new mechanisms of materials degradation are also possible. The purpose of the US Department of Energy Office of Nuclear Energy’s Light Water Reactor Sustainability (LWRS) Program is to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the operating lifetimes of nuclear power plants beyond 60 years. The intent of this LWRS Concrete nondestructive evaluation (NDE) roadmap is to define Research & Development (R&D) actions to address gaps between available NDE concrete techniques and the technology needed to make quantitative measurements to determine the durability and performance of concrete structures on our current NPP fleet. To assist in the development of this roadmap, a workshop (See Appendix B for Workshop Agenda) was held on July 31, 2012, at the Oak Ridge National Laboratory (ORNL) Conference Center with 28 attendees: three representatives from the Electrical Power Research Institute (EPRI), one representative from Electricité de France, two representatives from the US Nuclear Regulatory Commission (NRC), six representatives from universities (University of Tennessee-Knoxville, Georgia Tech, Iowa State University, University of Illinois, and University of Houston), one representative from industry (WesDyne), one representative from a non-USA nonprofit organization (Swiss Association for Technical Inspection), and 14 representatives from DOE national laboratories (ORNL, Pacific Northwest National Laboratory, Idaho National Laboratory, and Argonne National Laboratory). A multitude of concrete-based structures are typically part of a light water reactor (LWR) plant to provide foundation, support, shielding, and containment functions. Concrete has been used in the construction of nuclear power plants (NPPs) because of three primary properties—its inexpensiveness, its structural strength, and its ability to shield radiation. Examples of concrete structures important to the safety of LWR plants include the containment building, spent fuel pool, and cooling towers. This use has made its long-term performance crucial for the safe operation of commercial NPPs. With respect to the concrete structures, age-related degradation may affect engineering properties, structural resistance/capacity, failure mode, and location of failure initiation that in turn may affect the ability of a structure to withstand challenges in service. In order to ensure the safe operation of NPPs, it is essential that the effects of potential degradation of the plant structures, as well as systems and components, be assessed and managed during both the current operating license period as well as subsequent license renewal periods. In contrast to many mechanical and electrical components, replacement of many concrete structures is impractical. Therefore it is necessary that safety issues related to plant aging and continued service of the concrete structures are resolved through sound scientific and engineering understanding. Unlike most metallic materials, reinforced concrete is a nonhomogeneous material, a composite with low-density matrix, a mixture of cement, sand, aggregate and water, and a high-density reinforcement (typically 5% in NPP containment structures), made up of steel rebar or tendons. Plants have been typically built with local cement and aggregate fulfilling the design specification regarding strength, workability, and durability, but as a consequence, each plant’s concrete composition is unique and complex. In addition, NPP concrete structures are often inaccessible and contain large volumes of massively thick concrete. These structures are exposed to different environments (moisture, temperature) and a diversity of degradation mechanisms (high temperatures, radiation exposure, chemical reactions) at different plant sites, all of which adds to the complexity of determining the integrity/quality of the concrete.

Concrete↗
DOE OSTI2019

The corrosion of UK vitrified waste during disposal

The UK's vitrified high-level-waste (HLW) is destined to be disposed of in a deep geological disposal facility (GDF). Contact with groundwater will be the process by which radionuclides are released during 100,000+ years of disposal. Understanding the mechanisms and kinetics of glass dissolution allow models and predictions to be made on the long-term fate of the disposed HLW. Current standard laboratory scale experimental methodologies may be unrepresentative of the geological disposal environment. Aim: Provide insight into the dissolution behaviour of alkali-alumino-borosilicate UK vitrified waste in geological disposal environments. A 'cyclic vapor-static' dissolution methodology is proposed, capturing expected temporal behaviour. This will elucidate dissolution kinetics and mechanisms, optimize the base glass composition, waste loading content and generate significant, pragmatic disposal performance assessment data. Variants of three glasses to be investigated: waste-loaded MW25 and ZnCa MW28, and the ZnCa MW base. Preliminary Results: Suggests drying process decreases the chemical durability of the glass. Addition of ZnCa to UK HLW improves chemical durability. Increased waste-loading improves chemical durability. Banded structure in altered glass possibly attributed to dissolution/reprecipitation of secondary phases). Conclusion - Further Work: Cyclic vapor-static dissolution methodology may elucidate dissolution behaviour expected within a UK GDF. Potential to provide beneficial, pragmatic data for the performance assessment to implement geological disposal. Fundamental mechanisms to be understood.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Interdigital Capacitance Local Non-Destructive Examination of Nuclear Power Plant Cable for Aging Management Programs

This Pacific Northwest National Laboratory milestone report describes progress to date on the investigation of non-destructive test methods focusing on local cable insulation and jacket testing using an interdigital capacitance (IDC) approach. Earlier studies have assessed a number of non-destructive examination (NDE) methods for bulk, distributed, and local cable tests. A typical test strategy is to perform bulk assessments of the cable response using dielectric spectroscopy, Tan , or partial discharge followed by distributed tests like time domain reflectometry or frequency domain reflectometry to identify the most likely defect location followed by a local test that can include visual inspection, indenter modulus tests, or Fourier Transform Infrared Spectroscopy (FTIR) or Near Infrared Spectroscopy FTIR (FTNIR). If a cable is covered with an overlaying jacket, the jacket’s condition is likely to be more severely degraded than the underlying insulation. None of the above local test approaches can be used to evaluate insulation beneath a cable jacket. Since the jacket’s function is neither structural nor electrical, a degraded jacket may not have any significance regarding the cable’s performance or suitability for service. IDC measurements offer a promising alternative or complement to these local test approaches including the possibility to test insulation beneath an overlaying jacket.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Portland cement paste behaviour under gamma radiation field in geological repository - 15272

The Radioactive Waste Management Laboratory at the Nuclear and Energy Research Institute in Sao Paulo, Brazil (IPEN-CNEN-SP), is developing the concept of a repository for disposition of disused sealed radioactive sources in a deep borehole in which cement paste is intended to be used as a backfill between the steel casing and the geological formation around the borehole. The cement paste is intended to function as structural material, a blockage against the transport of water between the different strata of the geological setting, and as an additional barrier against the migration of radionuclides. The assessment of the cement paste behaviour in long term is necessary to increase the confidence that the material will perform as required during the service life of the facility. Under radiation, pore water present in cement paste suffers radiolysis and some products can be highly reactive, as electrons and hydroxyl radicals and hydrogen peroxide. These radiolysis products will interact with cement paste and its hydration products, forming a wide range of compounds. Previous work focused on assessing the synergic effects of radiation and environmental conditions, like higher temperature and aggressive groundwater chemicals, on cement paste. Examination of samples detected small differences in mineralogy and microstructure. However, the radiation dose to which samples were exposed (400 kGy) was only a fraction of the expected accumulated dose in cement paste until complete decay of some sealed sources in the repository. Therefore, several thousand kilo-grays must be tested the possibility of radiation to alter cement mineralogy and, as a consequence, its durability The objective of this research is to evaluate the changes in mineralogy and microstructure of irradiated cement paste specimens, in an attempt to establish a relationship between irradiation and durability. Cubic Portland cement specimens were casted and exposed to the gamma radiation of a 60 Co multipurpose irradiator, at IPEN-CNEN/SP. The effects of doses between 1000 kGy and 4000 kGy are evaluated by X-Ray Diffraction, Scanning Electron Microscopy, X-ray micro tomography, neutrongraphy and neutron scattering. This report presents the changes in cement paste mineralogy when samples are examined by XRD. Results indicate that durability can be affected because of the observed changes. Results are expected to indicate the induced changes in cement paste under high radiation doses and to provide hints on its behaviour in the long-term and service-life. The results can be used in safety analysis for deep geological repositories that use cement based materials as engineered barriers. (authors)

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

Neutron transfer reactions on the ground state and isomeric state of a 130 Sn beam

The structure of nuclei around the neutron-rich nucleus 132 Sn is of particular interest due to the vicinity of the Z = 50 and N = 82 shell closures and the r-process nucleosynthetic path. Four states in 131 Sn with a strong single-particle-like component have previously been studied via the (d,p) reaction, with limited excitation energy resolution. Here, the 130 Sn( 9 Be, 8 Be) 131 Sn and 130 Sn( 13 C, 12 C) 131 Sn single-neutron transfer reactions were performed in inverse kinematics at the Holifield Radioactive Ion Beam Facility using particle-γ coincidence spectroscopy. The uncertainties in the energies of the single-particle-like states have been reduced by more than an order of magnitude using the energies of γ rays. The previous tentative J π values have been confirmed. Decays from high-spin states in 131 Sn have been observed following transfer on the isomeric component of the 130 Sn beam. The improved energies and confirmed spin-parities of the p-wave states important to the r-process lead to direct-semidirect cross-sections for neutron capture on the ground state of 130 Sn at 30 keV that are in agreement with previous analyses. A similar assessment of the impact of neutron-transfer on the isomer would require significant nuclear structure and reaction theory input. There are few measurements of transfer reaction on isomers, and this is the first on an isomer in the 132 Sn region.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Dependencies of α' Embrittlement in Neutron-Irradiated Model Fe-Cr-Al Alloys

Fe-Cr-Al alloys are currently under consideration as accident tolerant fuel cladding materials due to their superior performance in high temperature steam environments when compared to Zr-based alloys. These alloys have also shown promise for other nuclear reactor and power production applications due to their excellent high-temperature oxidation and corrosion resistance. However, like other high-Cr ferritic Fe-based alloys, these alloys are prone to radiation-induced hardening and embrittlement due to the precipitation of a Cr-rich-α' phase under typical LWR operating conditions. Additionally, the kinetics of this phase separation have been shown to be accelerated by irradiation. Current research efforts are focused on developing an ideal candidate Fe-Cr-Al alloy composition and microstructure that retains its exceptional environmental compatibility while maintaining adequate mechanical performance over the duration of its operational lifetime. The current work investigates both the composition and dose dependencies of α' phase precipitation, as it can be the primary contributor to radiation-induced hardening and embrittlement in these alloys. The comprehensive analysis of phase stability in irradiated Fe-Cr-Al alloys provided by this work is a key step forward in assessing the radiation tolerance and overall viability of these materials for accident tolerant fuel cladding applications. Additional work is underway to assess radiation-induced dislocation loop formation in these specimens and correlate the morphologies of both the dislocation loop and α' precipitate morphologies to the results of mechanical testing campaigns in order to fully develop the structure-property relationships for this class of alloys. Completion of this work will allow for a comprehensive evaluation of the suitability of Fe-Cr-Al alloys for nuclear applications and result in the recommendation of an optimized candidate alloy ideally suited for LWR environments. (authors)

36 MATERIALS SCIENCE↗