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DOE OSTI2017

A HUMAN FACTORS ENGINEERING PROCESS TO SUPPORT HUMAN-SYSTEM INTERFACE DESIGN IN CONTROL ROOM MODERNIZATION

The primary objective of the United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is to sustain operation of the existing commercial nuclear power plants (NPPs) through a multi-pathway approach in conducting research and development (R&D). The Advanced Instrumentation, Information, and Control (II&C) System Technologies pathway conducts targeted R&D to address aging and reliability concerns with legacy instrumentation and control (I&C) and other information systems in existing U.S. NPPs. Control room modernization is an important part following this pathway, and human factors experts at Idaho National Laboratory (INL) have been involved in conducting R&D to support migration of new digital main control room (MCR) technologies from legacy analog and legacy digital I&C. This paper describes a human factors engineering (HFE) process that supports human-system interface (HSI) design in MCR modernization activities, particularly with migration of old digital to new digital I&C. The process described in this work is an expansion from the LWRS Report INL/EXT-16-38576, and is a requirements-driven approach that aligns with NUREG-0711 requirements. The work described builds upon the existing literature by adding more detail around key tasks and decisions to make when transitioning from HSI Design into Verification and Validation (V&V). The overall objective of this process is to inform HSI design and elicit specific, measurable, and achievable human factors criteria for new digital technologies. Upon following this process, utilities should have greater confidence with transitioning from HSI design into V&V.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2017

A Human Factors Engineering Process to Support Human-System Interface Design in Control Room Modernization

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2018

RISA Industry Use Case Analysis

This report provides a status of development of utility RISA Use Case applications for which potential sponsoring utility partners have been identified. These applications were developed with utility participation during meetings held at Idaho National Laboratory (INL) on 15 May 2018. Each of these Use Case applications meets the needs of both plant owner / operators and the objectives of the LWRS Program. These applications also meet the needs of the host utilities and the overall industry in that they can be accomplished within a timeframe that can provide economic value to both the sponsoring host utility and the current operating fleet. In particular, the methods and tools that will be developed during these Use Case applications are be designed with the intent that they can be in place and supporting near term resolution of the issue (i.e. within the next 5 years) across the industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Charpy Impact Characterization of Surveillance Specimens Harvested from Palisades High Fluence A-60 Capsule

Located on the shores of Lake Michigan, the Palisades Nuclear Generating Station (PNGS) was a nuclear power plant that operated in Covert Township, Michigan. The plant had a single pressurized water reactor that produced electricity for the region. The PNGS was shut down in 2022 after more than four decades of service. The PNGS included in its surveillance program a surveillance capsule, designated A-60, containing specimens of a weld metal with nickel content of about 1.36 wt% and copper content of about 0.25 wt%. The capsule was removed from its surveillance position in early 1995 and has been resident in the spent fuel pool since that time. This capsule was irradiated to a fluence of 1.96×10 20 n/cm 2 (E>1MeV) that is equivalent for more than 150 effective full power years (EFPYs) for the US reactor pressure vessel (RPV) fleet. The material is also of special interest because of its very high nickel content and potential for development of NiMnSi (nickel-manganese-silicon) precipitates. Combination of very high fluence and very high Ni and Cu content makes the material in this capsule of great interest as benchmark for currently developing embrittlement trend curves (ETC) aiming to predict embrittlement at high fluences. Multi-year efforts by the Light Water Reactor Sustainability (LWRS) Program personnel from the Materials Research Pathway (MRP) to harvest this capsule finally succeeded in 2023. As a result, the Westinghouse Electric Company (WEC) through contract with Oak Ridge National Laboratory (ORNL) came to PNGS site, retrieved the A-60 capsule, brought it to WEC Churchill hot cell facility, opened the capsule and sent all surveillance specimens in the capsule to ORNL for future characterization by July 2023. Total of 9 tensile and 48 Charpy specimens were inventoried in the ORNL hot cells. Testing plan has been developed based on available specimens. It includes hardness, tensile, Charpy impact, Mini-CT fracture toughness testing, in-situ thermal annealing, and microstructural characterization, including Atom Probe Tomography (APT). Charpy impact testing has been completed and results are presented in this report. Moreover, negotiations with Pressurized Water Reactors Owners Group (PWROG) and Westinghouse resulted in Westinghouse donating to ORNL a piece of archive weld and base metal such that unirradiated characterization of these materials can be performed as part of this project. Comparison of measured hardening and embrittlement has been performed and data are compared to available large database.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Human factors engineering process to support human-system interface design in control room modernization - 351

The primary objective of the United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is to sustain operation of the existing commercial nuclear power plants (NPPs) through a multi-pathway approach in conducting research and development (R and D). The Advanced Instrumentation, Information, and Control (II and C) System Technologies pathway conducts targeted R and D to address aging and reliability concerns with legacy instrumentation and control (I and C) and other information systems in existing U.S. NPPs. Control room modernization is an important part following this pathway, and human factors experts at Idaho National Laboratory (INL) have been involved in conducting R and D to support migration of new digital main control room (MCR) technologies from legacy analog and legacy digital I and C. This paper describes a human factors engineering (HFE) process that supports human-system interface (HSI) design in MCR modernization activities, particularly with migration of old digital to new digital I and C. The process described in this work is an expansion from the LWRS Report INL/EXT-16-38576, and is a requirements-driven approach that aligns with NUREG-0711 requirements. The work described builds upon the existing literature by adding more detail around key tasks and decisions to make when transitioning from HSI Design into Verification and Validation (V and V). The overall objective of this process is to inform HSI design and elicit specific, measurable, and achievable human factors criteria for new digital technologies. Upon following this process, utilities should have greater confidence with transitioning from HSI design into V and V. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Use of collaboration software to improve nuclear power plant outage management

Nuclear Power Plant (NPP) refueling outages create some of the most challenging activities the utilities face in both tracking and coordinating thousands of activities in a short period of time. Other challenges, including nuclear safety concerns arising from atypical system configurations and resource allocation issues, can create delays and schedule overruns, driving up outage costs. Today the majority of the outage communication is done using processes that do not take advantage of advances in modern technologies that enable enhanced communication, collaboration and information sharing. Some of the common practices include: runners that deliver paper-based requests for approval, radios, telephones, desktop computers, daily schedule printouts, and static whiteboards that are used to display information. Many gains have been made to reduce the challenges facing outage coordinators; however; new opportunities can be realized by utilizing modern technological advancements in communication and information tools that can enhance the collective situational awareness of plant personnel leading to improved decision-making. Ongoing research as part of the Light Water Reactor Sustainability Program (LWRS) has been targeting NPP outage improvement. As part of this research, various applications of collaborative software have been demonstrated through pilot project utility partnerships. Collaboration software can be utilized as part of the larger concept of Computer-Supported Cooperative Work (CSCW). Collaborative software can be used for emergent issue resolution, Outage Control Center (OCC) displays, and schedule monitoring. Use of collaboration software enables outage staff and subject matter experts (SMEs) to view and update critical outage information from any location on site or off.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

An Integrated Framework for Risk Assessment of Safety-related Digital Instrumentation and Control Systems in Nuclear Power Plants: Methodology Refinement and Exploration

This report documents activities performed by Idaho National Laboratory (INL) during Fiscal Year (FY) 2023 for the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) risk assessment project. In FY 2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a technical basis to support effective, and secure DI&C technologies for digital upgrades/designs. A risk assessment-informed framework was proposed for this strategy, which aims to (1) provide a best-estimate, risk informed capability to quantitatively estimate the safety margin obtained from plant modernization, especially for safety-related DI&C systems, (2) support and supplement existing risk informed DI&C design guides by providing quantitative risk information and evidence, (3) offer a capability of design architecture evaluation of various DI&C systems, (4) assure the long-term safety and reliability of safety-related DI&C systems, and (5) reduce uncertainty in costs and support integration of DI&C systems in the plant. To achieve these technical goals, the LWRS-developed framework provides a means to address relevant technical issues by: (1) defining a risk informed analysis process for DI&C upgrade that integrates hazard analysis, reliability analysis, and consequence analysis, (2) applying risk informed tools to address common cause failures (CCFs) and quantify corresponding failure probabilities for DI&C technologies, particularly software CCFs, (3) evaluating the impact of digital failures at the component level, system level, and plant level, and (4) providing insights and suggestions on designs to manage the risks, thus to support the development and deployment of advanced DI&C technologies in nuclear power plants (NPPs). Adding diversity within a system or components is the primary means to eliminate and mitigate CCFs, but diversity also increases system complexity and may not address all sources of systematic failures. Optimization of diversity and redundancy applications for the safety-critical DI&C systems remains a challenge. To deal with the technical issues in addressing potential software CCFs in safety-related DI&C systems of NPPs and supporting relevant design optimization, the proposed framework provides: (a) A best-estimate, risk informed capability to address new technical digital issues quantitatively, focusing on software CCFs in safety-related DI&C systems of NPPs; (b) A common and a modularized platform for DI&C designers, software developers, cybersecurity analysts, and plant engineers to predict and prevent risk in the early design stage of DI&C systems; (c) Technical bases and risk informed insights to assist users address the risk informed alternatives for evaluation of CCFs in safety-related DI&C systems of NPPs; and (d) A risk informed tool that offers a capability of design architecture evaluation of various DI&C systems to support system design decisions in diversity and redundancy applications. The research and development efforts of this project in FY 2023 are focused on refining current methods on software CCF modeling and estimation and exploring additional innovative approaches to risk assessment of DI&C systems to enable a more comprehensive and complete assessment of various safety-related DI&C design architectures. The primary audience of this report are DI&C designers, engineers, and probabilistic risk assessment (PRA) practitioners. This includes stakeholders, such as the nuclear utilities and regulators who consider the deployment and upgrade of DI&C systems, DI&C software developers and reviewers, and cybersecurity specialists. It should be noted that all the analyses are performed for the demonstration of the methodology, not for the evaluation of an actual digital control system. Results are obtained based on limited design information and testing data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Plant Modernization Pathway: FY 2019 External Review Summary Report

This report describes an external review conducted by the Light Water Reactor Sustainability Program Plant Modernization Pathway (referred to as Pathway) to solicit feedback on the topics and results of the ongoing Instrumentation and Control (I&C) research program. This review was held in conjunction with the Nuclear Energy Institute (NEI) Digital I&C Working Group meeting that was held at Idaho National Laboratory on October 16–17, 2018. Members of the Nuclear Regulatory Commission (NRC) staff working on digital I&C issues through their Integrated Action Plan also attended. The NEI and NRC agreed that the meeting participants would serve as an external review panel for the purpose of obtaining expert input on the value and timing of the research projects. Five major research areas were presented, along with demonstrations of new technologies in the Human Systems Simulation Laboratory and the Computer-Assisted Virtual Environment. Following the presentations, the external review was conducted by referring to a set of summary slides of the five research areas, and allowing each participant to enter ratings and comments on a survey form as each area was addressed. Fourteen forms were returned. These feedback forms were analyzed by the Pathway who compiled and responded to the data and comments, which are documented in this report. The participants also had opportunity to provide comments on the LWRS Program as a whole and to suggest additional research topics of value. The feedback provided by the external review participants is taken to be a strong endorsement of the types of projects being conducted by the Pathway, the value they hold for the operating nuclear plants, and the general timing of need. The feedback aligns well with the priorities, levels of efforts allocated for the research projects, and project schedules of the current Pathway research program. The feedback also provides realistic observations on the practicality of some aspects of implementing these technologies. The Pathway will consider all of the feedback and address the recommended research topics that have been identified by the external review participants. Pathway research plans will be adjusted as warranted.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Risk Analysis of Various Design Architectures for High Safety-significant Safety-related Digital Instrumentation and Control Systems of Nuclear Power Plants during Accident Scenarios

This report documents the plus-up activities performed by Idaho National Laboratory (INL) during Fiscal Year (FY) 2022 for the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) risk assessment project. In FY 2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a strong technical basis to support effective, licensable, and secure DI&C technologies for digital upgrades/designs. An integrated risk assessment technology for the DI&C systems was proposed for this strategy, which aims to (1) provide a best-estimate, risk-informed capability to quantitatively and accurately estimate the safety margin obtained from plant modernization, especially for the high safety-significant safety-related (HSSSR) DI&C systems, (2) support and supplement existing advanced risk-informed DI&C design guides by providing quantitative risk information and evidence, (3) offer a capability of design architecture evaluation of various DI&C systems to support system design decisions and diversity and redundancy applications, (4) assure the long-term safety and reliability of HSSSR DI&C systems, and (5) reduce uncertainty in costs and support integration of DI&C systems in the plant. To achieve these technical goals and deal with the expensive licensing justifications from regulatory insights, the LWRS-developed framework instructs nuclear vendors and utilities on how to effectively lower the costs associated with digital compliance and speed industry advances by: (1) defining an integrated risk-informed analysis process for DI&C upgrade, including hazard analysis, reliability analysis, and consequence analysis, (2) applying systematic and risk-informed tools to address common cause failures (CCFs) and quantify corresponding failure probabilities for DI&C technologies, particularly software CCFs, (3) evaluating the impact of digital failures at the component level, system level, and plant level, and (4) providing insights and suggestions on designs to manage the risks, thus to support the development, licensing, and deployment of advanced DI&C technologies on nuclear power plant (NPPs). Adding diversity within system or components is the main means to eliminate and mitigate CCFs, but diversity also increases plant complexity and errors and may not address all sources of systematic failures. How to optimize the diversity and redundancy applications for the safety-critical DI&C systems remains a challenge. To deal with the technical issues in addressing potential software CCFs in HSSSR DI&C systems of NPPs and supporting relevant design optimization, the framework provides: ? An integrated best-estimate, risk-informed capability to address new technical digital issues quantitatively, accurately, and efficiently in plan modernization progress, such as software CCFs in HSSSR DI&C systems of NPPs ? A common and a modularized platform for DI&C designers, software developers, cybersecurity analysts, and plant engineers to efficiently predict and prevent risk in the early design stage of DI&C systems ? Technical bases and risk-informed insights to assist U.S. Nuclear Regulatory Commission (NRC) and industry to address and fulfill the risk-informed alternatives for evaluation of CCFs in HSSSR DI&C systems of NPPs ? An integrated risk-informed tool that offers a capability of design architecture evaluation of various DI&C systems to support system design decisions in diversity and redundancy applications. The plus-up research and development efforts of this project in FY 2022 are focused on methodology improvement of software CCF modeling and estimation, prevention analysis, importance analysis and risk analysis of various design architectures of HSSSR DI&C systems. This work greatly enhances the capability of the LWRS-developed framework for the risk assessment and design optimization of safety-critical DI&C systems. It should be noted that all the analyses are performed for the demonstration of the LWRS-developed framework, not for the evaluation of relevant systems. Results are obtained based on very limited design information and testing data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Activation Energies of Harvested Boston Insulated Wire and Okonite Cable Materials

As part of the Cable Aging Task within the Material Aging and Degradation (MAaD) pathway of the DOE Light Water Reactor Sustainability (LWRS) program, ORNL is collaborating with Pacific Northwest National Laboratory (PNNL), the Electric Power Research Institute (EPRI), and the US Nuclear Regulatory Commission (NCR) to study cable aging mechanisms. Understanding cable aging mechanisms in cable nuclear power plants (NPPs) with needed information as they seek plant life extensions to 80 years of operations. This report summarizes activities in the characterization of harvested electrical I&C cables related to activation energy and their role in determining the remaining margin of time cable insulation in NPPs can be used. Activation energy is the amount of energy that is required to break chemical bonds in a polymer through an endothermic reaction. For I&C cables that are utilized in NPPs, activation energy is used with a time and temperature dependent cable property like elongation at break (EAB) and/or an electrical property like voltage withstand to determine whether a cable insulation material meets the performance requirements for the environmental conditions such as temperature, radiation, and humidity for a given plant location. Considering differences in activation energies that were observed in harvested cable jackets when compared to previous measurements, an assessment of the cable insulation activation energy in harvested I&C cables was performed given that the insulation performance ultimately determines that lifetime performance for a cable. This assessment consisted of the mechanical and chemical characterization of harvested Boston Insulated Wire (BIW) and Okonite cable insulations from Zion NPP from a series of thermal accelerated aging exposures between 140°C and 160°C. Cables were obtained as part of the LWRS Zion Harvesting Project in cooperation with Energy Solutions and the NRC. Activation energies of 1.10 eV and 0.61 eV were obtained for BIW and Okonite manufactured ethylene propylene rubber (EPR) insulation with individual chlorosulfonated polyethylene (CSPE) jackets respectively. When compared to previous measurements, the activation energy of BIW agreed well, while there was a sizeable gap for previous Okonite activation energies near 0.91 eV. The lack of agreement could be attributed to differences in raw materials used in a given insulation and/or jacket, differences in the thicknesses of the EPR insulation and CSPE individual jacket or the environmental conditions that the insulation from previous measurements experienced prior to characterization especially if cables were not stored at manufacturer recommended conditions. Comparison of harvested insulation compositions from multiple NPP sites and locations in a given NPP with respect to activation energy could provide some guidance on property sensitivity and insulation variations. To address the issue of material thickness, an initial assessment of the influence of the CPSE individual jacket was also done. This was motivated by the different combinations of EPR insulation and CSPE individual jacket thicknesses that were found in harvested cables and previous measurements. Accelerated thermal aging of BIW-manufactured EPR insulation with individual CSPE jackets removed did show that the CSPE layer affected insulation degradation, further comparison of the compositional make-up of the CSPE and EPR for both BIW and Okonite insulation, along with a limited amount of additional accelerated aging, is needed to increase the level of confidence regarding the influence of the CSPE.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Evaluation of the Benefits of ATF, FLEX, and Passive Cooling System for an Enhanced Resilient PWR Model

This report, along with report INL/EXT-19-53556, documents the activities performed by Idaho National Laboratory (INL) during the fiscal year (FY) 2019 for the DOE Light Water Reactor Sustainability (LWRS) Program, Risk-Informed System Analysis (RISA) Pathway, Enhanced Resilient Plant (ERP) Systems research. This report includes additional accident scenarios for risk-informed ATF analysis, and examines the risk impacts of FLEX and passive cooling system on nuclear power plants (NPPs) from additional perspectives. The same analysis process, risk analysis approaches, and analysis tools as in FY 2018 were used for near-term ATF cladding (i.e., Iron-Chromium-Aluminum [FeCrAl] cladding and Chromium [Cr]-coated cladding) designs under the small break loss-of-coolant accident (SBLOCA) scenarios, two types of general transient scenarios (locked rotor and turbine trip), and main steam line break (MSLB) scenarios. Benchmark calculations for ATF analysis using RELAP5-3D and MELCOR are also presented. In addition, based on the generic probabilistic risk assessment (PRA) model, the risk impacts of FLEX and the dynamic natural convection (DNC) system on NPPs are evaluated through a risk-informed decision-making process, i.e., significance determination process (SDP).

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2018

Plant-Level Scenario-Based Risk Analysis for Enhanced Resilient PWR – SBO and LBLOCA

This report documents the activities performed during the FY2018 for the DOE Light Water Reactor Sustainability (LWRS) Program, Risk-Informed System Analysis (RISA) Pathway, Enhanced Resilient Plant (ERP) Systems research. The purpose of the RISA Pathway research and development is to support plant owner-operator decisions with the aim to improve the economics, reliability, and maintain the high levels of safety of current nuclear power plants over periods of extended plant operations. The concept of ERP refers to the combinations of Accident Tolerant Fuel (ATF), optimal use of Diverse and Flexible Coping Strategy (FLEX), enhancements to plant components and systems, and the incorporation of augmented or new passive cooling systems, as well as improved fuel cycle efficiency. The objective of the ERP research effort is to use the RISA methods and toolkit in industry applications, including methods development and early demonstration of technologies, in order to enhance existing reactors safety features (both active and passive) and to substantially reduce operating costs through risk-informed approaches to plant design modifications to the plant and their characterization. There are two main focus areas in FY2018 for the ERP R&D efforts. One is to evaluate the risk impact brought by the ATF designs and FLEX in the selected accident scenarios. The other is to investigate various approaches to accomplish the ERP research objective, i.e., use RISA methods and toolkit to enhance existing reactors safety features and reduce plant operating costs.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2019

Exploring the Psychometrics of Common Post-Scenario Human Factors Questionnaires of Workload, Situation Awareness, and Perceived Difficulty

Abstract: The United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program has the primary mission of sustaining operation of the existing commercial nuclear power plants (NPPs) through a multi-pathway approach in conducting research and development (R&D). In this effort, the role of human factors engineering (HFE) has been an integral in ensuring new digital upgrades that are part of the overall main control room modernization strategy for various U.S. NPPs optimize human-system performance and do not introduce new human failure modes. In this process, testing and evaluation of concept (i.e., prototype) human machine interfaces (HMIs) are done in operator-in-the-loop studies at various stages of the HFE modification process, such as during design and verification and validation (V&V) [1]. During these studies, licensed operators go through select scenarios using a testbed, such as a full-scale simulator or at a NPP’s training simulator, to identify potential usability issues and provide general design feedback regarding the new digital HMIs. Human-system performance is measured and evaluated through a range of objective and self-report measures that capture aspects of plant performance, task performance, workload, situation awareness, and anthropometric and ergonomic concerns [2]. The use of post-scenario questionnaires is an acceptable method for gathering measures of workload and situation awareness. To this end, the NASA-Task Load Index (TLX) [e.g., 3, 4, & 5] and Situation Awareness Rating Technique (SART) [e.g., 6 & 7] are common standardized questionnaires used for evaluating workload and situation awareness, respectively. When combined, these questionnaires consist of a total of 15 questions, each using a different rating scale. Debriefs are also done through interviews with operators to gather design insight after each scenario. Since operator- and training simulator-time is limited, a concern during this post-scenario period of operator-in-the-loop studies is the limited time available to run scenarios, administer post-scenario questionnaires, and debrief within the allotted simulator time. To address this potential problem, this paper explores the psychometric characteristics of NASA-TLX, SART, and a common one-question post-scenario usability questionnaire (i.e., the single ease question – SEQ [8]) in effort to inform future development of a condensed questionnaire that captures aspects of workload and situation awareness. This exploratory work uses data from four different operator-in-the-loop studies completed under prior LWRS operator-in-the-loop studies. Findings from this paper should support future development of a consolidated questionnaire that may be used in operator-in-the-loop studies, when time is compressed, that conforms to the HFE scientific and regulatory requirements set for NPP control room modernization.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2016

Benchmark Accelerated Aging of Harvested Hypalon/Epr and Cspe/Xlpe Power and I&C Cable in Nuclear Power Plants

As part of the Light Water Reactor and Sustainability (LWRS) program in the U.S. Department of Energy (DOE) Office of Nuclear Energy, material aging and degradation research is currently geared to support the long-term operation of existing nuclear power plants (NPPs) as they move beyond their initial 40 year licenses. The goal of this research is to provide information so that NPPs can develop aging management programs (AMPs) to address replacement and monitoring needs as they look to operate for 20 years, and in some cases 40 years, beyond their initial operating lifetimes. For cable insulation and jacket materials that support instrument, control, and safety systems, accelerated aging data are needed to determine priorities in cable aging management programs. Before accelerated thermal and radiation aging of harvested, representative cable insulation and jacket materials, the benchmark performance of a new test capability at Oak Ridge National Laboratory (ORNL) was evaluated for temperatures between 70 and 135°C, dose rates between 100 and 500 Gy/h, and accumulated doses up to 20 kGy, Samples that were characterized and are representative of current materials in use were harvested from the Callaway NPP near Fulton, Missouri, and the San Onofre NPP north of San Diego, California. From the Callaway NPP, a multiconductor control rod cable manufactured by Boston Insulated Wire (BIW), with a Hypalon/ chorolosulfonated polyethylene (CSPE) jacket and ethylene-propylene rubber (EPR) insulation, was harvested from the auxiliary space during a planned outage in 2013. This cable was placed into service when the plant was started in 1984. From the San Onofre NPP, a Rockbestos Firewall III (FRIII) cable with a Hypalon/ CSPE jacket with cross-linked polyethylene (XLPE) insulation was harvested from an on-site, climate-controlled storage area. This conductor, which was never placed into service, was procured around 2007 in anticipation of future operation that did not occur. Benchmark aging for both jacket and insulation material was carried out in air at a temperature of 125°C or in uniform 140 Gy/h gamma irradiation over a period of 60 days. Their mechanical properties over the course of their exposures were compared with reference data from comparable cable jacket/insulation compositions and aging conditions. For both accelerated thermal and radiation aging, it was observed that the mechanical properties for the Callaway BIW control rod cable were consistent with those previously measured. However, for the San Onofre Rockbestos FRIII, there was an observable functional difference for accelerated thermal aging at 125°C. Details on possible sources for this difference and plans for resolving each source are given in this paper.

aging↗

DOE OSTI2015

Human Performance Modeling for Dynamic Human Reliability Analysis

Part of the U.S. Department of Energy’s (DOE’s) Light Water Reac- tor Sustainability (LWRS) Program, the Risk-Informed Safety Margin Charac- terization (RISMC) Pathway develops approaches to estimating and managing safety margins. RISMC simulations pair deterministic plant physics models with probabilistic risk models. As human interactions are an essential element of plant risk, it is necessary to integrate human actions into the RISMC risk framework. In this paper, we review simulation based and non simulation based human reliability analysis (HRA) methods. This paper summarizes the founda- tional information needed to develop a feasible approach to modeling human in- teractions in RISMC simulations.

human reliability analysis↗

DOE OSTI2016

In-air and pressurized water reactor environment fatigue experiments of 316 stainless steel to study the effect of environment on cyclic hardening

Argonne National Laboratory (ANL), under the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316SS) material which is widely used in the US reactors. Contrary to the conventional S~N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In a second paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.

Cyclic plasticity material modeling↗

DOE OSTI2014

Research in nondestructive evaluation techniques for nuclear reactor concrete structures

The purpose of the Materials Aging and Degradation (MAaD) Pathway of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program is to develop the scientific basis for understanding and predicting longterm environmental degradation behavior of material in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components (SSCs) essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on SSCs is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e. service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enable by improved methods and techniques for detection, monitoring, and prediction of SSC degradation. The MAaD Pathway R and D Roadmap for Concrete, 'Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap', focused initial research efforts on understanding the recent concrete issues at nuclear power plants and identifying the availability of concrete samples for NDE techniques evaluation and testing. [1] An overview of the research performed by ORNL in these two areas is presented here.

42 ENGINEERING↗

DOE OSTI2017

CURRENT STATUS OF THE CHARACTERIZATION OF RPV MATERIALS HARVESTED FROM THE DECOMMISSIONED ZION UNIT 1 NUCLEAR POWER PLANT

The decommissioning of Units 1 and 2 of the Zion Nuclear Power Station in Zion, Illinois, after ~ 15 effective full-power years of service presents a unique opportunity to characterize the degradation of in-service reactor pressure vessel (RPV) materials and to assess currently available models for predicting radiation embrittlement of RPV steels [1-3]. Moreover, through-wall thickness attenuation and property distributions are being obtained and the results to be compared with surveillance specimen test data. It is anticipated that these efforts will provide a better understanding of materials degradation associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service and subsequent license renewal. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the U.S. Department of Energy, Light Water Reactor Sustainability (LWRS) Program, coordinated procurement of materials, components, and other items of interest from the decommissioned Zion NPPs. In this report, harvesting, cutting sample blocks, machining test specimens, test plans, and the current status of materials characterization of the RPV from the decommissioned Zion NPP Unit 1 will be discussed. The primary foci are the circumferential, Linde 80 flux, wire heat 72105 (WF-70) beltline weld and the A533B base metal from the intermediate shell harvested from a region of peak fluence (0.7 X 1019 n/cm2, E > 1.0 MeV) on the internal surface of the Zion Unit 1 vessel. Following the determination of the through-thickness chemical composition, Charpy impact, fracture toughness, tensile, and hardness testing are being performed to characterize the through-thickness mechanical properties of base metal and beltline-weld materials. In addition to mechanical properties, microstructural characterizations are being performed using various microstructural techniques, including Atom Probe Tomography, Small Angle Neutron Scattering, and Positron Annihilation Spectroscopy

Rosseel, T. M.↗