Shah Malik

Damage Assessment Technologies for Prognostics and Proactive Management of Materials Degradation

Abstract The U.S. Nuclear Regulatory Commission has undertaken a program to lay the groundwork for defining proactive actions to manage degradation of materials in light water reactors (LWRs). This proactive management of materials degradation (PMMD) program examines LWR component materials and the degradation phenomena that affect them. Of particular interest is how such phenomena can be monitored and data can be used to predict degradation and prevent component failure. Some forms of degradation, including some modes of stress corrosion cracking, are characterized by a long initiation time followed by a rapid growth phase, and monitoring such long-term degradation will require new nondestructive evaluation methods and measurement procedures. As reactor lifetimes are extended, degradation mechanisms previously considered too long-term to be of consequence (such as concrete and wiring insulation degradation) may become more important. This paper explains the basic principles of PMMD and its relationship to in-service inspection, condition-based maintenance, and advanced diagnostics and prognostics. It then reviews the phases for degradation development and technologies with potential for sensing and monitoring degradation in its early stages.

An updated probabilistic fracture mechanics methodology for application to pressurized thermal shock

The US Nuclear Regulatory Commission (NRC) is, in concert with the US nuclear industry, currently revisiting its rule and analysis requirements for pressurized thermal shock (PTS) scenarios. This paper provides an overview of an updated probabilistic risk analysis (PRA) methodology that is continuing to evolve as part of that effort. The evolution process includes a careful assessment of recent advancements that have been made in the various parts of the computational methodologies. The process also involves interactions between experts in relevant disciplines (thermal hydraulics, PRA, materials, fracture mechanics, and non-destructive and destructive examination). Representatives include staff members from the USNRC staff, research laboratories, and the nuclear industry. The updated methodology is being integrated into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code for application to re-examine the adequacy of the current regulations and to determine if the updates provide sufficient technical bases for revisions. This paper also discusses recent modifications to the probabilistic fracture mechanics (PFM) methodology that is central to FAVOR.

Proactive Management of Materials Degradation for nuclear power plant systems

There are approximately 440 operating reactors in the global nuclear power plant (NPP) fleet that have an average age greater than 20 years and design lives of 30 or 40 years. The United States is currently implementing license extensions of 20 years on many plants, and consideration is now being given to the concept of ldquolife-beyond-60,rdquo a further period of license extension from 60 to 80 years and potentially longer. In almost all countries with NPPs, authorities are looking at some form of license renewal program. There is a growing urgency as a number of plants face either approvals for license renewal or shut down, which will require deployment of new power plants. In support of NPP license renewal over the past decade, various national and international programs have been initiated. This paper reports part of the work performed in support of the U.S. Nuclear Regulatory Commissionpsilas (NRCpsilas) Proactive Management of Materials Degradation (PMMD) program. The paper concisely explains the basic principles of PMMD and its relationship to advanced diagnostics and prognostics and provides an assessment of some the technical gaps in PMMD and prognostics that need to be addressed.

Historical Context of Elevated Temperature Structural Integrity for Next Generation Plants: Regulatory Safety Issues in Structural Design Criteria of ASME Section III Subsection NH

In 2006, ASME and DOE signed a cooperative agreement to update and expand appropriate materials, construction and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. The second task in this ASME/DOE Gen-IV Materials Project was to identify issues relevant to ASME Section III, Subsection NH, and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. The Nuclear Regulatory Commission (NRC) and Advisory Committee on Reactor Safeguards (ACRS) issues which were raised in 1983 in conjunction with the licensing of the Clinch River Breeder Reactor (CRBR) provide the best early indication of regulatory licensing issues for high temperature reactors. The approach to resolve the 25 identified elevated temperature structural integrity licensing issues was never implemented because Congress halted the construction of CRBR. This 1983 list provided the most definitive description of NRC elevated temperature structural integrity concerns. This paper presents both the results of the study by O’Donnell and Griffin [1] and a preliminary analysis by NRC staff of the earlier identified elevated temperature structural integrity issues that attempts to provide updated information for several of the next generation reactor types under consideration.Copyright

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (F racture A nalysis of V essels: O ak R idge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.Copyright

Analysis of Emerging NDE Techniques: Methods for Evaluating and Implementing Continuous Online Monitoring

There are approximately 440 operating reactors in the global nuclear power plant (NPP) fleet with an average age greater than 20 years and original design lives of 30 or 40 years. The United States is currently implementing license extensions of 20 years on many plants, and consideration is now being given to the concept of “life-beyond-60”, license extension from 60 to 80 years and potentially longer. In almost all countries with NPPs, authorities are looking at some form of license renewal program. In support of NPP license renewal over the past decade, various national and international programs have been initiated. One of the goals of the program for the proactive management of materials degradation (PMMD) is to manage proactively the in-service degradation of metallic components in aging NPPs. As some forms of degradation, such as stress corrosion cracking, are characterized by a long initiation time followed by a rapid growth phase, new inspection or monitoring technologies may be required. New nondestructive evaluation (NDE) techniques that may be needed include techniques to find stress corrosion cracking (SCC) precursors, on-line monitoring techniques to detect cracks as they initiate and grow, as well as advances in NDE technologies. This paper reports on the first part of the development of a methodology to determine the effectiveness of these emerging NDE techniques for managing metallic degradation. This methodology will draw from experience derived from evaluating techniques that have “emerged” in the past. The methodology will follow five stages: a definition of inspection parameters, a technical evaluation, laboratory testing, round robin testing, and the design of a performance demonstration program. This methodology will document the path taken for previous techniques and set a standardized course for future NDE techniques. This paper then applies the expert review section of the methodology to the acoustic emission technique to evaluate the use of acoustic emission in performing continuous online monitoring of reactor components.Copyright

AGING MANAGEMENT USING PROACTIVE MANAGEMENT OF MATERIALS DEGRADATION

The U.S. Nuclear Regulatory Commission (NRC) has undertaken a program to lay the technical foundations for defining proactive actions to manage degradation of materials in light water reactors. The current focus is existing plants; however, if applied to new construction, there is potential to better monitor and manage plants throughout their life cycle. This paper discusses the NRC’s Proactive Management of Materials Degradation program and its application to nuclear power plant structures, systems, and components.

USING TECHNOLOGY TO SUPPORT PROACTIVE MANAGEMENT OF MATERIALS DEGRADATION FOR THE U.S. NUCLEAR REGULATORY COMMISSION

The majority of the U.S. reactor fleet is applying for license renewal to extend the operating life from the current 40 years to 60 years, and there is now active interest in extending the operating life to beyond 60 years. Many plants are also applying for increases in power rating and both of these changes increases the need for an improved understanding of materials degradation. Many materials degrade over time and much is known about the degradation of materials under normal environmental conditions; however, less is known about the characteristics of materials degradation when the environment is subject to higher than normal radiological conditions over extended periods of time. Significant efforts are being made by industrial, academic and regulatory groups worldwide to identify, classify and mitigate potential problems arising from degradation of components in this context. From a regulatory perspective, the U.S. Nuclear Regulatory Commission (NRC) is very interested in being able to identify ways to ensure their licensees proactively manage the identification of materials degradation and the mitigation of its effects. To date, the NRC has consolidated “generic” programs for mitigating aging issues in the two volume Generic Aging Lessons Learned (GALL) Report (NUREG-1801) and has encouraged applicants for license renewal to use these programs where applicable in their plant when applying for renewal of their reactor’s license. The NRC has also published a comprehensive report entitled Expert Panel Report on Proactive Materials Degradation (NUREG/CR-6923) [3]. This report inventories the types of degradation mechanisms that could exist in each component of a light water reactor (LWR) and indicates how much is known about mitigating the effects within that context. Since the number of plant designs and materials used varies greatly within the U.S. fleet, there are many variations to implementing aging management programs (AMPs), requiring significant dialogs between the licensee and the NRC. These discussions are part of the licensing basis and as such are documented with up to multi-hundred page responses that are loosely coupled through the NRC Agency-wide Document Access and Management System (ADAMS). ADAMS serves as an electronic records repository for the NRC. These discussions have supported revisions to the GALL, including the revision that is being prepared as this paper is being written. The NRC has sought the help of the Pacific Northwest National Laboratory (PNNL) to improve their staff’s ability to navigate the significant numbers of documents that are generated in this process. PNNL is also to provide a forum for regulators, licensees, and researchers to share knowledge in their efforts to improve the cyclic process for defining, applying, validating, and re-defining AMPs. Work to date in this area is publicly accessible, and this paper will describe that work and outline a potential path forward. The presenter will also demonstrate the capabilities of the PMMD information tools (http://pmmd.pnl.gov).Copyright

Structural Integrity Code and Regulatory Issues in the Design of High Temperature Reactors

Since the 1980s, the ASME Code has made numerous improvements in elevated-temperature structural integrity technology. These advances have been incorporated into Section II, Section VIII, Code Cases, and particularly Subsection NH of Section III of the Code, “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400°F (760°C) to about 1742°F (950°C) where creep effects limit structural integrity, safe allowable operating conditions, and design life. Materials that are more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses. Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high temperature applications. Current ASME Section III design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking and corresponding criteria are needed for high temperature reactors. Subsection NH of Section III was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC, DOE and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. The design lifetime of Gen IV Reactors is expected to be 60 years. Additional materials including Alloy 617 and Hastelloy X need to be fully characterized. Environmental degradation effects, especially impure helium and those noted herein, need to be adequately considered. Since cyclic finite element creep analyses will be used to quantify creep rupture, creep fatigue, creep ratcheting and strain accumulations, creep behavior models and constitutive relations are needed for cyclic creep loading. Such strain- and time-hardening models must account for the interaction between the time-independent and time-dependent material response. This paper describes the evolving structural integrity evaluation approach for high temperature reactors. Evaluation methods are discussed, including simplified analysis methods, detailed analyses of localized areas, and validation needs. Regulatory issues including weldment cracking, notch weakening, creep fatigue/creep rupture damage interactions, and materials property representations for cyclic creep behavior are also covered.Copyright