The Sensitivity of Pressurized Thermal Shock Analysis Results to Alternative Models for Weld Flaw Distributions

2003 ◽  
Author(s):  
T. L. Dickson ◽  
F. A. Simonen
Keyword(s):  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
Evaluation Study ◽  
The United States ◽  
Metal Plate ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Pressurized Thermal Shock ◽  
Technical Basis ◽  
The Impact

The United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project in 1999 to determine if improved technologies can provide a technical basis to reduce the conservatism in the current regulations for pressurized thermal shock (PTS) 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 renewal considerations. During the PTS re-evaluation study, an improved risk-informed computational methodology was developed that provides a more realistic characterization of PTS risk. This updated methodology was recently applied to three commercial PWRs. The results of this study provide encouragement that a technical basis can be established to support a relaxation of current PTS regulations. One significant model improvement applied in the PTS re-evaluation study was the development of flaw databases derived from the non-destructive and destructive examinations of material from cancelled reactor pressure vessels (RPV). Empirically-based statistical distributions derived from these databases and expert illicitation were used to postulate the number, size, and location of flaws in welded and base metal (plate and forging) regions of an RPV during probabilistic fracture mechanics (PFM) analyses of RPVs subjected to transient loading conditions such as PTS. However, limitations in the available flaw data have required assumptions to be made to complete the risk-based flaw models. Sensitivity analyses were performed to evaluate the impact of four flaw-related assumptions. Analyses addressed: 1) truncations of distributions to exclude flaws of extreme depth dimensions, 2) vessel-to-vessel differences in flaw data, 3) large flaws observed in weld repair regions, and 4) the basis for estimating the number of surface breaking flaws. None of the four alternate weld flaw models significantly impacted calculated vessel failure frequencies or invalidated the tentative conclusions derived from the USNRC PTS re-evaluation study.

The Impact of an Improved Flaw Model on a Pressurized Thermal Shock Evaluation

2002 ◽  
Author(s):  
T. L. Dickson ◽  
F. A. Simonen
Keyword(s):  
Thermal Shock ◽  
Computational Models ◽  
The United States ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Evaluation Program ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
The Impact ◽  
Flaw Characterization

The current regulations for pressurized thermal shock (PTS) were derived from computational models that were developed in the early-mid 1980s. The computational models utilized in the 1980s conservatively postulated that all fabrication flaws in reactor pressure vessels (RPVs) were inner-surface breaking flaws. It was recognized at that time that flaw-related data had the greatest level of uncertainty of the inputs required for the probabilistic-based PTS evaluations. To reduce this uncertainty, the United States Nuclear Regulatory Commission (USNRC) has in the past few years supported research at Pacific Northwest National Laboratory (PNNL) to perform extensive nondestructive and destructive examination of actual RPV materials. Such measurements have been used to characterize the number, size, and location of flaws in various types of welds and the base metal used to fabricate RPVs. The USNRC initiated a comprehensive project in 1999 to re-evaluate the current PTS regulations. The objective of the PTS Re-evaluation program has been to incorporate advancements and refinements in relevant technologies (associated with the physics of PTS events) that have been developed since the current regulations were derived. There have been significant improvements in the computational models for thermal hydraulics, probabilistic risk assessment (PRA), human reliability analysis (HRA), materials embrittlement effects on fracture toughness, and fracture mechanics methodology. However, the single largest advancement has been the development of a technical basis for the characterization of fabrication-induced flaws. The USNRC PTS-Revaluation program is ongoing and is expected to be completed in 2002. As part of the PTS Re-evaluation program, the updated risk-informed computational methodology as implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, including the improved PNNL flaw characterization, was recently applied to a domestic commercial pressurized water reactor (PWR). The objective of this paper is to apply the same updated computational methodology to the same PWR, except utilizing the 1980s flaw model, to isolate the impact of the improved PNNL flaw characterization on the PTS analysis results. For this particular PWR, the improved PNNL flaw characterization significantly reduced the frequency of RPV failure, i.e., by between one and two orders of magnitude.

Derivation of the New Pressurized Thermal Shock Screening Criteria

2011 ◽  
Author(s):  
Terry Dickson ◽  
Shengjun Yin ◽  
Mark Kirk ◽  
Hsuing-Wei Chou
Keyword(s):  
United States ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Technical Basis

As a result of a multi-year, multi-disciplinary effort on the part of the United States Nuclear Regulatory Commission (USNRC), its contractors, and the nuclear industry, a technical basis has been established to support a risk-informed revision to pressurized thermal shock (PTS) regulations originally promulgated in the mid-1980s. The revised regulations provide alternative (optional) reference-temperature (RT)-based screening criteria, which is codified in 10 CFR 50.61(a). How the revised screening criteria were determined from the results of the probabilistic fracture mechanics (PFM) analyses will be discussed in this paper.

An Overview of the Pressurized Thermal Shock Re-Evaluation Project

2005 ◽  
Author(s):  
Terry L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Thermal Shock ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Technical Basis ◽  
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.

Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During Pressurized Thermal Shock Events

2013 ◽  
Vol 136 (1) ◽  
Author(s):  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Crack Initiation ◽  
Thermal Shock ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Conditional Probabilities ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

To apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), a PFM analysis code has been developed at JAEA. Using this PFM analysis code, pascal version 3, the conditional probabilities of crack initiation (CPIs) and fracture for an RPV during pressurized thermal shock (PTS) events have been analyzed. Sensitivity analyses on certain input parameters were performed to clarify their effect on the conditional fracture probability. Comparisons between the conditional probabilities and the temperature margin (ΔTm) based on the current deterministic analysis method were made for various model plant conditions for typical domestic older types of RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.

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

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

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 (Fracture Analysis of Vessels: Oak Ridge) 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.

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

2008 ◽  
Author(s):  
Terry Dickson ◽  
Mark EricksonKirk
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Inner Surface ◽  
Regulatory Commission ◽  
Technical Basis

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.

Scoping Analyses of Parameterized Cool-Down Transients Associated With Reactor Shutdown

2007 ◽  
Author(s):  
T. L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Operating Temperature ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Operating Conditions ◽  
Temperature And Pressure ◽  
Reactor Operating ◽  
Technical Basis

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 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. During the past decade, the NRC conducted the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project that established a technical basis to support a risk-informed revision to current PTS regulations (10CFR Part 50.61). Once the results of the PTS reevaluation are incorporated into a revision of the 10 CFR 50 guidance on PTS, it is anticipated that the regulatory requirements for the fracture toughness of the RPV required to withstand a PTS event (accidental loading) will in some cases be less restrictive than the current requirements of Appendix G to 10 CFR Part 50, which apply to normal operating conditions. This logical inconsistency occurs because the new PTS guidelines will be based on realistic models and inputs whereas existing Appendix G requirements contain known and substantial conservatisms. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements of Appendix G to 10 CFR Part 50 in a manner that is consistent with that used to develop the risk-informed revision to the PTS regulations. Scoping probabilistic fracture mechanics (PFM) analyses have been performed for several hundred parameterized cool-down transients to (1) obtain insights regarding the interaction of operating temperature and pressure parameters on the conditional probability of crack initiation and vessel failure and (2) determine the limits on the permissible combinations of operating temperature and pressure within which the reactor may be brought into or out of an operational condition that remains below the acceptance criteria adopted for PTS of 1 × 10−6 failed RPVs per reactor operating year. This paper discusses the modeling assumptions, results, and implications of these scoping analyses.

Risk-Based Fracture Evaluation of Reactor Vessels Subjected to Cool-Down Transients Associated With Shutdown: An Examination of the Effects of Different Modeling Approaches on Estimated Failure Probabilities

2006 ◽  
Author(s):  
T. L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Asme Code ◽  
Computational Methodology ◽  
Regulatory Commission ◽  
Technical Basis ◽  
Evaluation Project

The current regulations, as set forth by the United States Nuclear Regulatory Commission (USNRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned 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. In 1999, the USNRC initiated the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project to determine if a technical basis could be established to support a relaxation in the current PTS regulations. The PTS re-evaluation project included the development and application of an updated risk-based computational methodology that incorporates several advancements applicable to modeling the physics of vessel fracture due to thermal hydraulic transients imposed on the RPV inner surface. The results of the PTS re-evaluation project demonstrated that there is a sound technical basis to support a relaxation of the current PTS regulations. The results of the PTS re-evaluation are currently under review by the USNRC. Based on the promising results of the PTS re-evaluation, the USNRC has recently applied the updated computational methodology to fracture evaluations of RPVs subjected to planned cool-down transients, associated with reactor shutdown, derived in accordance with ASME Section XI – Appendix G. The objective of these analyses is to determine if a sound technical basis can be established to provide a relaxation to the current regulations for the derivation of bounding cool-down transients as specified in Appendix G to Section XI of the ASME Code. This paper provides a brief overview of these analyses, results, and the implications of the results.

Alternate Requirements for Protection Against Pressurized Thermal Shock (PTS)

2013 ◽  
Author(s):  
Stephen M. Parker ◽  
Nathan A. Palm ◽  
Xavier Pitoiset
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Prediction Techniques ◽  
The U.S

Plants in the United States (U.S.) and many plants outside of the U.S. are required to meet the regulations of the Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61. The Alternate Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61a) was approved by the U.S. Nuclear Regulatory Commission (NRC) and included in the Federal Register, with an effective date of February 3, 2010. This Alternate Rule provides a new metric and screening criteria for PTS. This metric, RTMAX-X, and the corresponding screening criteria are far less restrictive than the RTPTS metrics and screening criteria in the original PTS Rule (10 CFR 50.61). The Alternate PTS Rule was developed through probabilistic fracture mechanics (PFM) evaluations performed for selected U.S. pilot plants. A Generalization Study was also performed which determined that the plants used for these evaluations were representative of and applicable to the U.S. Pressurized Water Reactor (PWR) nuclear power plant fleet. Plants outside of the U.S. may be interested in implementing the Alternate PTS Rule. However, direct implementation of the Alternate PTS Rule may not be possible due to differences in plant design, embrittlement prediction techniques, inservice inspection requirements, etc. The objective of this paper is to explore the use the Alternate PTS Rule by PWR plants outside of the U.S. by proposing methods to account for the potential differences mentioned above.

scholarly journals Survival in elderly glioblastoma patients treated with bevacizumab-based regimens in the United States

2018 ◽  
Vol 5 (4) ◽  
pp. 251-261 ◽  
Author(s):  
Jessica Davies ◽  
Irmarie Reyes-Rivera ◽  
Thirupathi Pattipaka ◽  
Stephen Skirboll ◽  
Beatrice Ugiliweneza ◽  
...  
Keyword(s):  
Elderly Patients ◽  
Lower Risk ◽  
Karnofsky Performance Status ◽  
The United States ◽  
Cox Proportional Hazards Model ◽  
Sensitivity Analyses ◽  
Primary Analysis ◽  
Risk Of Death ◽  
Initial Surgery ◽  
The Impact

AbstractBackgroundThe efficacy of bevacizumab (BEV) in elderly patients with glioblastoma remains unclear. We evaluated the effect of BEV on survival in this patient population using the Survival, Epidemiology, and End Results (SEER)-Medicare database.MethodsThis retrospective, cohort study analyzed SEER-Medicare data for patients (aged ≥66 years) diagnosed with glioblastoma from 2006 to 2011. Two cohorts were constructed: one comprised patients who had received BEV (BEV cohort); the other comprised patients who had received any anticancer treatment other than BEV (NBEV cohort). The primary analysis used a multivariate Cox proportional hazards model to compare overall survival in the BEV and NBEV cohorts with initiation of BEV as a time-dependent variable, adjusting for potential confounders (age, gender, Charlson comorbidity index, region, race, radiotherapy after initial surgery, and diagnosis of coronary artery disease). Sensitivity analyses were conducted using landmark survival, propensity score modeling, and the impact of poor Karnofsky Performance Status.ResultsWe identified 2603 patients (BEV, n = 597; NBEV, n = 2006). In the BEV cohort, most patients were Caucasian males and were younger with fewer comorbidities and more initial resections. In the primary analysis, the BEV cohort showed a lower risk of death compared with the NBEV cohort (hazard ratio, 0.80; 95% confidence interval, 0.72–0.89; P < .01). The survival benefit of BEV appeared independent of the number of temozolomide cycles or frontline treatment with radiotherapy and temozolomide.ConclusionBEV exposure was associated with a lower risk of death, providing evidence that there might be a potential benefit of BEV in elderly patients with glioblastoma.

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