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).

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.

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.

Severe Transportation Accidents: Do Used Nuclear Fuel Transportation Packages Survive Real World Accidents?

2011 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton ◽  
Harold Adkins ◽  
Judith Cuta ◽  
Nicholas Klymyshyn ◽  
...  
Keyword(s):  
Pacific Northwest ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Analysis Model ◽  
Element Analysis ◽  
Tanker Truck ◽  
The U.S

In 2007, a severe transportation accident occurred near Oakland, California, on a section of Interstate 880 known as the “MacArthur Maze,” involving a gasoline tanker truck which impacted an overpass support column and burst into flames. The fire caused the collapse of portions of the Interstate 580 overpass onto the remains of the tractor-trailer. The U.S. Nuclear Regulatory Commission, with assistance from Pacific Northwest National Laboratory, the Center for Nuclear Waste Regulatory Analyses, the Southwest Research Institute, and the National Institute of Standards and Technology, examined the accident conditions in order to characterize the fire and collapse that occurred, analyzed material samples from the collapsed I-580 overpass as well as the gasoline tanker truck, and developed a fire model of the accident. This was followed by development of a finite element analysis model to determine the impacts of this accident on the thermal and structural performance of a spent nuclear fuel (SNF) transportation package. The analysis results will be used to determine any potential regulatory implications related to the safe transport of SNF in the U.S. This paper provides a summary of this effort and presents some preliminary results and conclusions.

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.

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