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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Safety-Related Instrumentation and Control Upgrade Pilot Project: NUREG- 0711 Process Development, Planning and Analysis Activities, and Lessons Learned

Constellation Energy and the United States Department of Energy (DOE) have established a public/private partnership to implement a pilot digital upgrade to replace legacy analog, safety-related reactor protection and emergency safety feature actuation systems (RPS/ESFAS) with modern digital systems. This effort is occurring at Constellation’s Limerick Generating Station. This project is being performed in accordance with industry processes that have been adapted to better support digital upgrades. These processes include IP-ENG-001, Standard Design Process, NISP-EN-04, Standard Digital Engineering Process and Electric Power Research Institute Report 3002011816, Digital Engineering Guide. The Light Water Reactor Sustainability (LWRS) Program at the Idaho National Laboratory (INL) has been supporting this effort. Latest INL HFE efforts in support of this project have focused on definition and implementation of the Human Factors Engineering (HFE) Program in support of Constellation design and related licensing efforts. This paper presents the development execution of the HFE Program through completion of the Planning and Analysis Phase as defined by NUREG-0711 as well as HFE lessons learned.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

INTEGRATED RISK ASSESSMENT OF DIGITAL I&C SAFETY SYSTEMS FOR NUCLEAR POWER PLANTS

Upgrading the existing analog instrumentation and control (I&C) systems to state-of-the-art digital I&C (DI&C) systems provides the foremost means to improve performance and reduce costs for existing light-water reactors (LWRs). However, qualification of digital technologies remains a challenge?especially the issue of software common cause failure (CCF), which has been difficult to address. Existing analyses of CCFs in I&C systems mainly focus on hardware failures. With the application and upgrading of new DI&C systems, software CCFs due to design flaws might become a potential threat to plant safety, considering that most redundancy designs use similar digital platforms or software in their operating and application systems. With complex multi-layer redundancy designs to meet the single failure criterion, these I&C safety systems are of particular concern in U.S. Nuclear Regulatory Commission (NRC) licensing procedures. In 2019, the Risk-Informed Systems Analysis (RISA) Pathway of the U.S. Department of Energy?s (DOE?s) Light Water Reactor Sustainability (LWRS) Program initiated a project to develop a risk assessment strategy for delivering a strong technical basis to support effective, licensable, secure DI&C technologies for digital upgrades/designs. An integrated risk assessment for the DI&C (RADIC) process was proposed for this strategy to identify potential key digital-induced failures, implement reliability analyses of related digital safety I&C systems, and evaluate the unanalyzed sequences introduced by these failures (particularly software CCFs) at the plant level. This paper summarizes these RISA efforts in the risk analysis of safety-related DI& systems at Idaho National Laboratory.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

NEUTRON-IRRADIATION–INDUCED DAMAGE ASSESSMENT IN CONCRETE USING COMBINED PHASE CHARACTERIZATION AND NONLINEAR FAST FOURIER TRANSFORM SIMULATION

Understanding irradiation effects in concrete cavities of light water reactors is critical to the long-term operation (60+ years) of US nuclear power plants. Susceptibility of concrete to neutron irradiation greatly varies as a function of its constituents: coarse aggregates, sand, and hardened cement paste (hcp). In particular, higher irradiation–susceptibility of aggregates was found to be a direct function of radiation-induced volumetric expansion (RIVE), or the propensity of aggregates to swell as a function of mineral contents, structures and textures. Irradiation-induced amorphization is accompanied by significant swelling, especially in silicates. For example, the maximum volumetric expansions of quartz and feldspars have been shown to be as large as ≈ 18% and ≈ 8%, respectively, while the change of density in calcite remains rather low (≈ 0.3–0.5%). High swelling is associated with a high number of covalent bonds (Si–O) and a high degree of polymerization of [SiO4] tetrahedrons. Differential RIVEs between rock-forming minerals cause high stresses and may potentially develop cracking within the aggregates at the interface between aggregates and the hcp and within the hcp. Modeling neutron-irradiation effects in concrete requires addressing the problem at a scale that captures both the details of the aggregates’ mineralogy and texture, which is <~ 100 μm, and the representative size of concrete, which is ~ 10 cm. This requires high-resolution, large mapping capabilities. To this end, the US Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program developed an original tool suite called Microstructure Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) in which Nonlinear 2D-fast Fourier transform (FFT) techniques are employed to model mechanical or irradiation-induced cracking of high-resolution complex microstructures derived from combined mapping techniques, including micro x-ray fluorescence (m-XRF), energy-dispersive x-ray spectroscopy (EDS), electron backscatter diffraction (EBSD), ellipsometry, and scanning electron microscopy (SEM). Irradiated mineral properties are derived from the Irradiated Minerals, Aggregate and Concrete (IMAC) database, which includes available relevant literature data. Rigorous coupled damage-creep simulations are obtained through sequential linear analysis. Illustrations of actual accelerated irradiation-damage propagation in a complex microstructures are presented herein.

Le Pape, Yann↗
DOE OSTI2012

Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

EBSD Characterization of Tubular Cladding Plasticity after Bend Testing: A Feasibility Study

The Light Water Reactor Sustainability (LWRS) Program developed a bend test method for tubular samples during FY12. The purpose of this report is to provide details of the Electron Backscattered Diffraction (EBSD) measurements performed on the various sample sets to provide a comparative baseline of the impact that the silicon carbide ceramic matrix composites (SiC-CMC) sleeves had on the deformation and plasticity properties of the Zircaloy-4 (Zr-4) cladding tube.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

RELAP-7: Demonstrating Seven-Equation, Two-Phase

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory. RELAP-7 will become the main reactor systems toolkit for the Risk-Informed Safety Margin Characterization Pathway of the Light Water Reactor Sustainability (LWRS) Program and the next generation tool in the RELAP reactor safety/systems analysis application series (i.e., the replacement for RELAP5). The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (i.e., the Multi-Physics Object-Oriented Simulation Environment).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Deployment and Overview of RAVEN Capabilities for a Probabilistic Risk Assessment Demo for a PWR Station Blackout

Starting in 2012, Idaho National Laboratory (INL) has been engaged to provide tools and methodologies to support a new approach to safety assessment and management of safety related issues for Nuclear Power Plants (NPPs). This activity is placed under the Risk Informed Safety Margin Characterization (RISMC) [2] umbrella that is one of the technical paths of the Department Of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program. RISMC takes a broader approach to risk evaluation by generating more comprehensive information. This information is used to gain a more accurate estimation of risk margins and to highlight feasible improvements that seek the optimal point between risk mitigation and economical viability of the nuclear energy option. This is, of course, done without taking any judgmental approach with respect to risk, which is a regulatory body task, and the economical viability, which is up to the single states and the open electricity market. RISMC is therefore composed of methodologies and tools. The augmented information that RISMC seeks to generate, as enabling decision tool, is not holistic with respect all possible metrics and operational aspects of a power plant; it is rather channeled toward the improvement of safety and the associated economical impact. While these are indeed two very broad metrics, they still allow a focus within the development of the underlying tools that, in a resource constrained environment (DOE founding, computational, mathematical) maximize the benefit of the program to the nuclear industry.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2013

Evaluation of Ultrasonic Techniques on Concrete Structures

Materials issues are a key concern for the existing nuclear reactor fleet as material degradation can increase maintenance, downtime, and risk. Extending reactor life to 60 years and beyond will likely increase the reactor’s susceptibility to, and the severity of known forms of degradation. Additionally, new mechanisms of materials degradation are also possible. The purpose of the U.S. 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 (NPPs) beyond 60 years.

Concrete↗
DOE OSTI2013

RELAP-7: Demonstrating the integration of two-phase flow components for an ideal BWR loop

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory. RELAP-7 will become the main reactor systems toolkit for the Risk-Informed Safety Margin Characterization Pathway of the Light Water Reactor Sustainability (LWRS) Program and the next generation tool in the RELAP reactor safety/systems analysis application series (i.e., the replacement for RELAP5). The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). During Fiscal-Year 2013, a number of physical components have been developed to support the demonstration calculations presented in this report. The case selected for demonstration of an idealized BWR loop in RELAP-7 includes the major components for the primary system of a BWR. The BWR loop has been successfully simulated to steady state with the RELAP-7 code. The next major stage of development is to demonstrate two-phase flow modeling capability through a simplified boiling water reactor station blackout analysis, which will be reported in the subsequent demonstration simulation report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Integrating Safety Assessment Methods using the Risk Informed Safety Margins Characterization (RISMC) Approach

Safety is central to the design, licensing, operation, and economics of nuclear power plants (NPPs). As the current light water reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of systems, structures, and components (SSC) degradations or failures that initiate safety significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated primarily based on engineering judgment backed by a set of conservative engineering calculations. The ability to better characterize and quantify safety margin is important to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development (R&D) in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, readability, and safety, the RISMC Pathway provides methods and tools that enable mitigation options known as margins management strategies. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. As the lead Department of Energy (DOE) Laboratory for this Pathway, the Idaho National Laboratory (INL) is tasked with developing and deploying methods and tools that support the quantification and management of safety margin and uncertainty.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable quantitative assessment of the CASL modeling of Crud-Induced Power Shift (CIPS) phenomenon, in particular, and the CASL advanced predictive capabilities, in general. This report is prepared for the Department of Energy’s Consortium for Advanced Simulation of LWRs program’s VUQ Focus Area.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

Post-irradiation Examination Plan for ORNL and University of California Santa Barbara Assessment of UCSB ATR-2 Irradiation Experiment

New and existing databases will be combined to support development of physically based models of transition temperature shifts (TTS) for high fluence-low flux (φ < 10 11 n/cm 2 -s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1×10 20 n/cm 2 (>1 MeV). All references to neutron flux and fluence in this report are for fast neutrons (>1 MeV). The reactor pressure vessel (RPV) task of the Light Water Reactor Sustainability (LWRS) Program is working with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural features from reactor surveillance materials with those from similar materials irradiated under high flux conditions in test reactors

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

GRIZZLY/FAVOR Interface Project Report

As part of the Light Water Reactor Sustainability (LWRS) Program, the objective of the GRIZZLY/FAVOR Interface project is to create the capability to apply GRIZZLY 3-D finite element (thermal and stress) analysis results as input to FAVOR probabilistic fracture mechanics (PFM) analyses. The one benefit of FAVOR to Grizzly is the PROBABILISTIC capability. This document describes the implementation of the GRIZZLY/FAVOR Interface, the preliminary verification and tests results and a user guide that provides detailed step-by-step instructions to run the program.

Dickson, Terry L.↗
DOE OSTI2014

Preliminary Conceptual Design of a Thick Concrete Nondestructive Evaluation Specimen

Many important safety structures in a nuclear power plant (NPP) are constructed using concrete. The purpose of the U.S. 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 NPPs beyond 60 years. In order to evaluate the current status of NPP structures, concrete inspection techniques must be developed and tested. Since in-service containment structures do not allow for destructive measures necessary to validate the accuracy of these inspection techniques, comparative testing of the various nondestructive evaluation (NDE) concrete measurement techniques will require concrete specimens with known material properties, voids, internal microstructure flaws, and reinforcement locations. This report examines the need and availability of large concrete specimens to allow comparative testing of various NDE techniques, and the structural characteristics necessary for the specimens to serve as relevant representations of actual concrete structures in commercial NPPs.

concrete NDE↗
DOE OSTI2014

Radiation Testing of a Low Voltage Silicone Nuclear Power Plant Cable

This report summarizes the results generated in FY13 for cable insulation in support of DOE's Light Water Reactor Sustainability (LWRS) Program, in collaboration with the US-Argentine Binational Energy Working Group (BEWG). A silicone (SiR) cable, which was stored in benign conditions for ~30 years, was obtained from Comision Nacional de Energia Atomica (CNEA) in Argentina. Physical property testing was performed on the as-received cable. This cable was artificially aged to assess behavior with additional analysis. SNL observed appreciable tensile elongation values for all cable insulations received, indicative of good mechanical performance. Of particular note, the work presented here provides correlations between measured tensile elongation and other physical properties that may be potentially leveraged as a form of condition monitoring (CM) for actual service cables. It is recognized at this point that the polymer aging community is still lacking the number and types of field returned materials that are desired, but SNL—along with the help of others—is continuing to work towards that goal. This work is an initial study that should be complimented with location-mapping of environmental conditions of CNEA plant conditions (dose and temperature) as well as retrieval, analysis, and comparison with in-service cables.

White II, Gregory Von↗
DOE OSTI2014

Radiation Testing of a Low Voltage Silicone Nuclear Power Plant Cable.

This report summarizes the results generated in FY13 for cable insulation in support of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program, in collaboration with the US-Argentine Binational Energy Working Group (BEWG). A silicone (SiR) cable, which was stored in benign conditions for %7E30 years, was obtained from Comision Nacional de Energia Atomica (CNEA) in Argentina with the approval of NA-SA (Nucleoelectrica Argentina Sociedad Anonima). Physical property testing was performed on the as-received cable. This cable was artificially aged to assess behavior with additional analysis. SNL observed appreciable tensile elongation values for all cable insulations received, indicative of good mechanical performance. Of particular note, the work presented here provides correlations between measured tensile elongation and other physical properties that may be potentially leveraged as a form of condition monitoring (CM) for actual service cables. It is recognized at this point that the polymer aging community is still lacking the number and types of field returned materials that are desired, but Sandia National Laboratories (SNL) -- along with the help of others -- is continuing to work towards that goal. This work is an initial study that should be complimented with location-mapping of environmental conditions of Argentinean plant conditions (dose and temperature) as well as retrieval, analysis, and comparison with in- service cables.

White, II, Gregory Von↗
DOE OSTI2014

Case Study for Enhanced Accident Tolerance Design Changes

The ability to better characterize and quantify safety margin is important to improved decision making about Light Water Reactor (LWR) design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant systems, structures, and components (SSCs) needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, reliability, and safety, the Risk Informed Safety Margin Characterization (RISMC) Pathway provides methods and tools that enable mitigation options known as risk informed margins management (RIMM) strategies.

97 MATHEMATICS AND COMPUTING↗