Results of a Virtual Round Robin Study to Estimate Probability of Detection for Dissimilar Metal Welds

2021 ◽  
Author(s):  
Ryan M. Meyer ◽  
Aimee E. Holmes ◽  
Romarie Morales ◽  
Iikka Virkkunen ◽  
Thiago Seuaciuc-Osorio ◽  
...  
Keyword(s):  
Pacific Northwest ◽  
Nuclear Power ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Round Robin ◽  
Probability Of Detection ◽  
Pod Analysis ◽  
Regulatory Commission

Abstract This paper presents efforts to overcome challenges with empirical probability of detection (POD) estimations in the nuclear power industry through the utilization of a novel virtual flaw method. A virtual round robin (VRR) study was conducted under the Program for Investigation Of NDE by International Collaboration (PIONIC), organized by the United States Nuclear Regulatory Commission (NRC) utilizing data generated by the virtual flaw method. Analysis of results from the VRR was performed by teams from Pacific Northwest National Laboratory (PNNL), Electric Power Research Institute (EPRI), and Aalto University. Empirically derived POD estimations are presented, and challenges associated with obtaining these estimations are discussed. The virtual flaw method is introduced and some details of its implementation for the VRR activity are described. Results from POD analysis of the VRR data by PNNL, EPRI, and Aalto University are presented and a discussion regarding differences in analysis results is provided. Finally, potential future efforts to improve the application of the virtual flaw method and its estimation of POD are discussed.

Enhanced VVER-1000 Fuel Technology and Performance

2009 ◽  
Author(s):  
H. Shah ◽  
R. Latorre ◽  
G. Raspopin ◽  
J. Sparrow
Keyword(s):  
Pacific Northwest ◽  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Safety Level ◽  
And Performance

The United States Department of Energy, through the Pacific Northwest National Laboratory (PNNL), provides management and technical support for the International Nuclear Safety Program (INSP) to improve the safety level of VVER-1000 nuclear power plants in Central and Eastern Europe.

Benchmarking of a Thermal Finite Element Approximation Scheme for Externally Cooled Spent Fuel Storage Casks

2007 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Ian F. Spivack
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Finite Element Approximation ◽  
Nuclear Regulatory Commission ◽  
Element Approximation ◽  
Spent Fuel ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Regulatory Commission

The US Nuclear Regulatory Commission (NRC) is responsible for licensing spent fuel storage casks under Title 10 of the Code of Federal Regulations Part 72 (10 CFR Part 72). Under these regulations, storage casks must be evaluated to verify that they meet various criteria, including acceptable thermal performance requirements. The purpose of the evaluation described in this paper is to establish the effectiveness of a medium-effort modeling approach and associated simplifying assumptions in closely approximating spent fuel cask component temperature distributions. This predictive evaluation is performed with the ANSYS® code, and is applicable to externally cooled cask designs. The results are compared against experimental measurements and predictions of the COBRA-SFS finite-difference code developed at Pacific Northwest National Laboratory.

Structural Analysis Approach for the Defense Programs Package 3 (DPP-3)

2020 ◽  
Author(s):  
Peter J. Sakalaukus ◽  
Nathan P. Barrett ◽  
Brian J. Koeppel
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Analytical Techniques ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Design Criteria ◽  
Element Analysis ◽  
Regulatory Testing ◽  
Regulatory Commission

Abstract The Pacific Northwest National Laboratory (PNNL) is the design authority for a new Type B hazardous materials transportation package designated as the Defense Programs Package 3 (DPP-3) for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA). The DPP-3 has been developed using similar materials and fabrication methods employed in previous U.S. Nuclear Regulatory Commission (NRC), DOE, and NNSA certified packages. The DPP-3 design criteria are derived from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), NNSA guidance and NRC regulatory guides in order to safely and securely transport a variety of payloads. Final regulatory approval by the NNSA will require regulatory testing to demonstrate that the containment vessel (CV) remains leaktight after enduring the entire regulatory testing sequence prescribed in Title 10 of the Code of Federal Regulations Part 71 (10 CFR 71). In order to gain confidence that the DPP-3 will remain leaktight after testing, the DPP-3 has been structurally analyzed using the Finite Element Analysis (FEA) software LS-DYNA. The FEA analyses serve two general purposes: first, they aid in design and development of the package, and second, they advise as to which drop orientations are expected to cause the most damage during regulatory testing. This paper will discuss how the design criteria are incorporated into analytical techniques needed to evaluate the FEA structural simulation results for 10 CFR 71 conditions to give confidence the DPP-3 testing campaign will be successful.

Distributions of Fabrication Flaws in Reactor Pressure Vessels for Structural Integrity Evaluations

2002 ◽  
Author(s):  
F. A. Simonen ◽  
G. J. Schuster ◽  
S. R. Doctor ◽  
T. L. Dickson
Keyword(s):  
Pacific Northwest ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Distribution Functions ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Reactor Pressure

To reduce uncertainties in flaw-related inputs for probabilistic fracture mechanics (PFM) evaluations, the U.S. Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) involving nondestructive and destructive examinations for fabrication flaws in reactor pressure vessel (RPV) material. Using these data, statistical distributions have been developed to characterize the flaws in regions of a RPV. The regions include the main seam welds, repair welds, base metal, and the cladding at the inner surface of the vessel. This paper summarizes the available data and describes the treatment of these data to estimate flaw densities, flaw depth distributions, and flaw aspect ratio distributions. The methodology has generated flaw-related inputs for PFM calculations that have been part of an effort to update pressurized thermal shock (PTS) regulations. Statistical treatments of uncertainties in the parameters of the flaw distribution functions are part of the inputs to the PFM calculations. The paper concludes with a presentation of some example input files that have supported evaluations by USNRC of the risk of vessel failures caused by PTS events.

NDE Assessment of PWSCC in Control Rod Drive Mechanism Housings

2006 ◽  
Author(s):  
Steven R. Doctor ◽  
Stephen E. Cumblidge ◽  
George J. Schuster ◽  
Robert V. Harris ◽  
Susan L. Crawford
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Primary Objective ◽  
Control Rod ◽  
Drive Mechanism ◽  
Nondestructive Examination ◽  
Service Inspection

Studies being conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington are focused on assessing the effectiveness of nondestructive examination (NDE) techniques for inspecting control rod drive mechanism (CRDM) nozzles and J-groove weldments. The primary objective of this work is to provide information to the United States Nuclear Regulatory Commission (US NRC) on the effectiveness of NDE methods as related to the in-service inspection of CRDM nozzles and J-groove weldments, and to enhance the knowledge base of primary water stress corrosion cracking (PWSCC) through destructive characterization of the CRDM assemblies. In describing two CRDM assemblies removed from service, decontaminated, and then used in a series of NDE measurements, this paper will address the following questions: 1) What did each technique detect?, 2) What did each technique miss?, and 3) How accurately did each technique characterize the detected flaws? Two CRDM assemblies including the CRDM nozzle, the J-groove weld, buttering, and a portion of the ferritic head material were selected for this study. One contained suspected PWSCC, based on in-service inspection data and through-wall leakage; the other contained evidence suggesting through-wall leakage, but this was unconfirmed. The two CRDMs used in this study were cut from a pressure vessel head that has since been replaced. The selected NDE measurements follow standard industry techniques for conducting in-service inspections of CRDM nozzles and the crown of the J-groove welds and buttering. In addition, laboratory based NDE methods were employed to conduct inspections of the CRDM assemblies, with particular emphasis on inspecting the J-groove weld and buttering. This paper will also describe the NDE methods used and discuss the NDE results. Future work will involve using the results from these NDE studies to guide the development of a destructive characterization plan to reveal the crack morphology and a comparison of the degradation found by the destructive evaluation with the recorded NDE responses.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

Validation of Reanalysis-Based Offshore Wind Resource Characterization Using Lidar Buoy Observations

2020 ◽  
Vol 54 (6) ◽  
pp. 44-61
Author(s):  
Lindsay M. Sheridan ◽  
Raghavendra Krishnamurthy ◽  
Alicia M. Gorton ◽  
Will J. Shaw ◽  
Rob K. Newsom
Keyword(s):  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
Offshore Wind ◽  
Approximation Techniques ◽  
The North ◽  
Wind Resource

AbstractThe offshore wind industry in the United States is gaining strong momentum to achieve sustainable energy goals, and the need for observations to provide resource characterization and model validation is greater than ever. Pacific Northwest National Laboratory (PNNL) operates two lidar buoys for the U.S. Department of Energy (DOE) in order to collect hub height wind data and associated meteorological and oceanographic information near the surface in areas of interest for offshore wind development. This work evaluates the performance of commonly used reanalysis products and spatial approximation techniques using lidar buoy observations off the coast of New Jersey and Virginia, USA. Reanalysis products are essential tools for setting performance expectations and quantifying the wind resource variability at a given site. Long-term accurate observations at typical wind turbine hub heights have been lacking at offshore locations. Using wind speed observations from both lidar buoy deployments, biases and degrees of correspondence for the Modern-Era Retrospective Analysis for Research and Applications-2 (MERRA-2), the North American Regional Reanalysis (NARR), ERA5, and the analysis system of the Rapid Refresh (RAP) are examined both at hub height and near the surface. Results provide insights on the performance and uncertainty of using reanalysis products for long-term wind resource characterization. A slow bias is seen across the reanalyses at both deployment sites. Bias magnitudes near the surface are on the order of 0.5 m s−1 greater than their hub height counterparts. RAP and ERA5 produce the highest correlations with the observations, around 0.9, followed by MERRA-2 and NARR.

Human Factor Data in Nuclear Power Plant Applications

1980 ◽  
Vol 24 (1) ◽  
pp. 123-123
Author(s):  
Linda O. Hecht
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Reliability Modeling ◽  
Plant Safety ◽  
Modeling Techniques ◽  
Nuclear Systems ◽  
Factor Data ◽  
Regulatory Commission ◽  
The Impact

Due to the concern for safety the nuclear power industry in the United States has fostered the use of reliability analysis to assess system performance and the impact of system failure on overall plant safety. The need for system and component failure rate data has been recognized and has spurred such efforts as NPRDS (Nuclear Power Research Data System) and IEEE's Std 500 (The Reliability Data Manual). Reliability modeling techniques have been developed for application to nuclear systems and are presently being considered by the Nuclear Regulatory Commission for licensing purposes.

An Ultrasonic Phased Array Evaluation of Intergranular Stress Corrosion Crack (IGSCC) Detection in Austenitic Stainless Steel Piping Welds

2010 ◽  
Author(s):  
Aaron A. Diaz ◽  
Michael T. Anderson ◽  
Anthony D. Cinson ◽  
Susan L. Crawford ◽  
Stephen E. Cumblidge
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Intergranular Stress Corrosion ◽  
Water Reactor ◽  
Nondestructive Examination ◽  
Steel Piping

Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light water reactor (LWR) components and challenging material/component configurations. This study assessed the effectiveness of far-side inspections on wrought stainless steel piping with austenitic welds, as found in thin-walled, boiling water reactor (BWR) component configurations, for the detection and characterization of intergranular stress corrosion cracks (IGSCC).

Incorporation of Risk Insights in the Regulatory Treatment of Nuclear Power Plant Structures, Systems, and Components

2002 ◽  
Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Special Treatment ◽  
Regulatory Commission ◽  
High Level

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.

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