Mark T. EricksonKirk

Insights and Observations Arising From Curve-Fitting the Charpy V-Notch and Tensile Data Contained Within the United States’ Light Water Reactor Surveillance Database

The effect of irradiation damage on the mechanical properties of reactor pressure vessel structures is monitored in operating nuclear reactors according to the provisions of 10 CFR Part 50 Appendix H. In these surveillance programs Charpy V-notch energy and tensile data are collected. Trends in these data have and continue to be used to identify and quantify embrittlement trends, which is a key aspect to maintaining the continued structural integrity of the operating reactor fleet. This paper presents the current results from an on-going investigation aimed at assessing the effect of different curve-fitting strategies on the insights that can be gained from these data trending activities.Copyright

Insights Arising From a Comparison of the tanh and Exponential Fitting Methods for Charpy V-Notch Energy Data

In the 1960s and 1970s when the surveillance programs for currently operating commercial nuclear reactors were established state of knowledge limitations resulted in the use of Charpy-V notch (CVN) specimens rather than fracture toughness specimens. Reasonable success has since been achieved in correlating CVN and fracture toughness parameters. Such correlations provide an important part of the technical basis for both current regulations and ASME codes. These correlations imply that trends manifest in CVN data must also appear in fracture toughness data even though empirical evidence demonstrates that this is not always true. For example, the temperature dependence of CVN energy (CVE) in transition is thought to be a unique feature of each specific sample of ferritic steel that is tested, a view in sharp contrast with the now widely accepted view of a “Master Curve” for transition fracture toughness (KJc ). Also, effects of product form on CVE temperature dependence and property correlations are widely reported despite the fact that product form effects are absent from KJc properties. These observations suggest that the mapping of CVE behavior onto fracture toughness implicit to correlation-based regulations and ASME codes may produce erroneous trends in estimated values of fracture toughness. In this paper we investigate the hypothesis that the apparent differences between CVE and fracture toughness arise due to differences in how the temperature dependence of CVE and KJc data have historically been modeled. Our analysis shows that when CVE data are analyzed in a manner consistent with KJc data (i.e., transition and upper shelf data are partitioned from each other and analyzed separately rather than being fit with a continuous tanh function) the apparent differences between CVE and toughness characterizations are minimized significantly, and may disappear entirely. These findings demonstrate the differences between CVE and fracture toughness data to be an artifact of the tanh analysis method rather than an intrinsic property of CVE.

The Sensitivity of Risk-Informed Reactor Structural Integrity Analysis Results to Various Interpretations of Warm Pre-Stress

Warm pre-stress, or WPS, is a phenomenon by which the apparent fracture toughness of ferritic steel can be elevated in the fracture mode transition if crack is first “pre-stressed” at an elevated temperature. Taking proper account of WPS is important to the accurate modeling of the postulated accident scenarios that, collectively, are referred to as pressurized thermal shock, and to the accurate modeling of routine cool-down transients. For both accident and routine cool-downs the transients begin at the reactor operating temperature (approximately 290°C for pressurized water reactors in the United States) and proceed to colder temperatures as time advances. The probabilistic fracture mechanics code FAVOR, which is being used by the NRC to provide the technical basis for risk-informed revisions of 10 CFR 50.61 and 10 CFR 50 Appendix G, adopts a model of WPS as part of its fracture driving force module. In this paper we assess the conservatism inherent to the FAVOR WPS model relative to a best-estimate WPS model constructed using data recently produced by the European Commission “SMILE” project and published by Moinereau and colleagues. Assessments of the conservatisms inherent to the so-called “conservative principle” WPS model, and also to a classic LEFM model that does not credit WPS are also made. The data presented herein demonstrate that, for an integrated analysis of PTS risk, the FAVOR and conservative principle WPS models both over-estimate the vessel failure risk by a factor of between 2 and 3× relative to the best estimate model. Our examination of the effect of WPS models on the predictions of individual transients reveals that for the severe transients that dominate risk there is little difference (usually less than 4×) between the conditional probabilities of crack initiation and of through wall cracking predicted by the different WPS models. There are considerable differences in the predictions of the various WPS and non-WPS models for low severity transients, however, the contribution of these transients to the total risk of vessel failure is small.Copyright

Use of Large Databases to Identify Trends in the Behavior of Ferritic Steels

In this paper we explore the crucial role played by the use of large databases in the identification, development, and refinement of models that describe the toughness behavior of ferritic steels. Specifically, we seek to emphasize and illustrate the point that when physical models are calibrated using large databases this process can reveal trends not previously seen, or foreseen. In support of this idea two examples are cited. First, the evidence for a CVE master curve in fracture mode transition is reviewed, as a counterpoint to the commonly held belief that each Charpy tanh transition curve is unique, with little commonality even within specific alloys, let alone across all ferritic steels. Second, new evidence is presented that the degree of prior hardening experienced by a ferritic steel has a systematic effect on the scatter exhibited by KJc data. This evidence suggests that the KJc Master Curve model, in which the scatter of KJc follows a Weibull distribution having a Kmin = 20 MPa√m and a slope (scatter magnitude) of 4, requires refinement, especially for the higher To values characteristic of steels that have been hardened by, for example, neutron irradiation damage.Copyright

The Inclusion of Inner Surface Breaking Flaws in Probabilistic Fracture Mechanics Analyses of Reactor Vessels Subjected to Planned Normal Cool-Down Transients

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements that reduces the conservatism and also is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). Previous publications have been successful in illustrating potential methods to provide a risk-informed relaxation to the current regulations for normal transients. Thus far, probabilistic fracture mechanics (PFM) analyses have been performed at 60 effective full power years (EFPY) for one of the reactors evaluated as part of the PTS re-evaluation project. In these previous analyses / publications, consistent with the assumptions utilized for this particular reactor in the PTS re-evaluation, all flaws for this reactor were postulated to be embedded. The objective of this paper is to review the analysis results and conclusions from previous publications on this subject and to attempt to modify / generalize these conclusions to include RPVs postulated to contain only inner-surface breaking flaws or a combination of embedded flaws and inner-surface breaking flaws.Copyright

Effect of Embrittlement Trend Curve Model on the Reference Temperature Based Screening Limits for Pressurized Thermal Shock

Recently, the USNRC has proposed a voluntary alternative to the pressurized thermal shock (PTS) rule (10CFR50.61(a)) that may be used instead of the existing PTS rule (10CFR50.61) by power reactor licensees to demonstrate the safety of their RPVs. The reference temperature (RT) screening limits in 10CFR50.61(a) are less restrictive than those of 10CFR50.61; this relaxation being justified by the much more realistic models that underlie the technical basis of the revised rule. One of the many models that constitute this technical basis is used to estimate the effect of irradiation damage on the fracture toughness transition temperature. This relationship, which is called the embrittlement trend curve, or ETC, can change as time goes on due to the continued development of new knowledge (e.g., from on-going reactor vessel surveillance programs, from test reactor studies aimed at better understanding the physical mechanisms of irradiation damage, etc.). In this paper we present information that quantifies the effect of different ETCs on the through wall cracking frequencies that are calculated using the probabilistic fracture mechanics computer code that was used to establish the RT limits that have been proposed in 10CFR50.61(a).Copyright

An Overview of the Pressurized Thermal Shock Re-Evaluation Project

In 1999, a study sponsored by the United States Nuclear Regulatory Commission (NRC) suggested that advances in the technologies associated with the physics of pressurized-thermal-shock (PTS) events developed since the derivation of the PTS regulations (established in the early-mid eighties) had the potential to establish a technical basis that could justify a relaxation in the current PTS-related regulations. A relaxation of these regulations could have profound implications for plant license extension considerations. Subsequently, the NRC initiated the interdisciplinary PTS Re-evaluation Project. During the five year project, an updated comprehensive computational methodology evolved, within the framework established by modern probabilistic risk assessment (PRA) techniques, through interactions among experts in relevant disciplines from the NRC staff, their contractors, and representatives from the nuclear industry. During 2004, the updated computational methodology was applied to three domestic commercial pressurized water reactors (PWRs). The most recent results of the PTS Re-evaluation Project provide a technical basis to support a relaxation of the current PTS regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. The details of the updated computational methodology, the mathematical models, the analysis results, key findings, and supporting information have recently been drafted in several very detailed and lengthy formal reports. These reports are currently under review at the NRC. An objective of this paper is to provide a short overview of the improved computational methodology, analysis results, and key findings of the PTS re-evaluation project. To demonstrate that a technical basis has been established to support a relaxation of the current PTS regulations, it is helpful to understand the derivation of the current PTS regulations; therefore, another objective of this paper is to contrast the interpretation of the analysis results of the PTS re-evaluation to those performed in the eighties from which the current PTS regulations were derived.Copyright