News & Views, Volume 53 | Encoded Phased Array Ultrasonic Examination Services for Cast Austenitic Stainless Steel (CASS) Piping Welds

/in , , , , , /by

IN PRESSURIZED WATER REACTOR (PWR) COOLANT SYSTEMS

By:  John Hayden and Jason Van Velsor

The CASS piping welds present in many PWR plants provide numerous and complicated challenges to their effective ultrasonic examinations. To this point, a viable ultrasonic examination solution for the inspection of these piping components, as required by ASME Code Section IX,  had previously not been available. By leveraging our technical expertise in materials, technology development, and advanced NDE deployment, Structural Integrity Associates, Inc (SI) has developed a new system that will provide a meaningful solution for the examination of CASS piping components. The result of this program will be the first commercial offering for the volumetric examination of CASS components in the nuclear industry.

BACKGROUND INFORMATION
ASME Section XI Class 1 RCS piping system welds fabricated using CASS materials pose serious and well-understood challenges to their effective ultrasonic examination. For decades, utilities and regulators have struggled with the administrative and financial burdens of Relief Requests, which were, and still are, based on the inability to perform meaningful volumetric examinations of welds in CASS components. 

Many years of futility and frustration may have fostered the belief that technology allowing effective and meaningful examination of CASS materials would never be achievable. This is no longer the case.

The failure mechanism for CASS material occurs through the loss of fracture toughness due to thermal aging embrittlement. The susceptibility of CASS material to thermal aging embrittlement is strongly affected by several factors, primary of which are system operating time and temperature, the casting method used during component manufacture, and molybdenum and ferrite content. In addition to the existing ASME Section XI requirements for the examination of welds in CASS materials, the susceptibility to thermal aging embrittlement drives the requirement for additional examinations (including ultrasonic examinations) as directed by several NRC-published NUREGs required for plant license renewal. The existence of a viable, effective examination capability for CASS materials plays a very important part in both currently required Inservice Inspections (ISI) and plant license renewal.

Figure 1. An example of the widely-varying microstructure of a centrifugally cast piping segment. False-color imaging is used to aid visualizing grain variations. (Image from NUREG/CR-6933 PNNL-16292)

CASS MATERIAL PROPERTIES AND EFFECT ON ULTRASONIC EXAMINATION
Metallurgical studies have revealed that the microstructure of CASS piping can vary drastically in the radial (through-wall) direction, as well as around the circumference and along the length of any given piping segment. Large and small equiaxed, columnar and mixed (combinations of equiaxed and columnar grains), and banding (layers of substantially different grain structures) are commonly observed in CASS piping materials. None of these conditions favor the performance of effective ultrasonic examinations.

Figure 2. PWR RCS Major Components

The very large and widely varying types (equiaxed, columnar, and randomly mixed), sizes and orientations of the anisotropic grains in CASS material are very problematic. Anisotropic is defined as an object or substance having a physical property that has a different value when measured in different directions. Such physical properties strongly affect the propagation of ultrasound in CASS material by causing severe attenuation (loss of energy through beam scattering and absorption), beam redirection, and unpredictable changes in ultrasonic wave velocity. These factors are responsible for the inability of ultrasonic examination to completely and reliably interrogate the Code-required volume (inner 1/3 Tnom) of welds in CASS piping material. Interestingly, CASS materials less than 1.6” Tnom (Pressurizer Surge Piping) can be effectively examined, while CASS materials over 2.00” (Main RCS Coolant Loop Piping) are less effectively examined.  Consequently, an ASME Section XI, Appendix VIII qualification program for CASS piping components has not been established and remains in the course of preparation. Nonetheless, ASME Section XI requirements to conduct inservice examinations of RCS piping welds fabricated from CASS components remain fully in force.

ASME CODE ACTIONS AFFECTING CASS PIPING EXAMINATIONS
ASME Section XI Code Case N-824, “Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface,” was approved by ASME in October 2012 and by the NRC in October 2019. This Code Case provides the first approved direction for the ultrasonic examination of welds joining CASS piping components. The ASME B&PV Code, Section XI, 2015 Edition, incorporates Code Case N 824 into Mandatory Appendix III in the form of Mandatory Supplement 2. To date, these two ASME Section XI Code documents remain the sole sources approved by ASME and NRC that provide specific direction for the examination of CASS RCS piping system welds and, therefore, form the foundation of SI’s approach for the development of our CASS ultrasonic examination solution.

SI’S CASS PROGRAM DESCRIPTION
SI is developing the industry’s most well-conceived and capable ultrasonic system for the examination of welds in CASS piping components. To accomplish this objective, SI has drawn upon our internal knowledge and experience, supplemented by a careful study of numerous authoritative bodies of knowledge relating to the examination of CASS components. The development of the SI examination system has been guided by both SI’s industry-leading 17 years of experience conducting phased array examinations in nuclear power plants and the knowledge acquired through the careful study of the topical information contained within industry-recognized publications. These published results of extensive industry research provided both guidance for the selection of phased array system components and CASS-specific material insights that strengthen the technical content of our Appendix III-based procedure. 

Figure 3. RCS Coolant Pump and Crossover Piping

CASS PROGRAM ELEMENTS
SI believes that the procedure, equipment and personnel featured in this program will be equivalent or superior to those that will form the industry-consensus approach for CASS ultrasonic examinations needed to successfully achieve Appendix VIII, (future) Supplement 9, “Qualification Requirements for Cast Austenitic Piping Welds.”

Ultrasonic Procedure – SI has crafted an ultrasonic examination procedure framework that is fully compliant with ASME Section XI, Mandatory Appendix III, Supplement 2, along with referenced Section XI Appendices as modified by the applicable regulatory documents.

Ultrasonic Equipment – SI has acquired and assembled the ultrasonic system components required by Code Case N-824 and Appendix III, Supplement 2, which includes the following:

  • Ultrasonic instrumentation capable of functioning over the entire expected range of examination frequencies. The standard examination frequency range extends from low-frequency, 500 KHz operation for RCS main loop piping welds through 1.0 MHz for pressurizer surge piping. 

SI has designed and acquired additional phased array transducers that meet the physical requirements of frequency, wave mode, and aperture size and are capable of generating the prescribed examination angles with the required focal properties. SI has designed and fabricated an assortment of wedge assemblies that will be mated with our phased array probes to provide effective sound field coupling to the CASS components being examined. SI’s wedge designs consider the CASS pipe outside diameter and thickness dimensions and employ natural wedge-to-material refraction to assure optimal energy transmission and sound field focusing.

SI also possesses several data encoding options that are necessary to acquire ultrasonic data over the expected range of component access and surface conditions. The encoding options will include:

  • Fully-automated scanning system, capable of driving the relatively large and heavy 500KHz phased array probes
  • The SI-developed Latitude manually-driven encoding system, which has been deployed during PDI-qualified dissimilar metal DM weld examinations in nuclear power plants

    Figure 4. Steam Generator Details

Examination Personnel – SI’s ultrasonic examination personnel are thoroughly trained and experienced in all elements of encoded ultrasonic data acquisition and analysis in nuclear plants. SI’s examiners have a minimum of 10 years of experience and hold multiple PDI qualifications in manual and encoded techniques. SI recognizes the challenges that exist with the examination of CASS piping welds and has developed a comprehensive program of specialized, mandatory training for personnel involved with CASS examinations. This training includes descriptions of coarse grain structures, their effect on the ultrasonic beam, and the expected ultrasonic response characteristics of metallurgical and flaw reflectors, as well as the evaluation of CASS component surface conditions.

ULTRASONIC TECHNQUE VALIDATION
Although not required by the ASME Code, SI has arranged for access to CASS piping system specimens from reputable sources to validate the efficiency of our data acquisition process and the performance of our ultrasonic examination techniques. The specimens represent various pipe sizes and wall thicknesses and contain flaws of known location and size to permit the validation and optimization of SI’s data acquisition and analysis processes. SI will thoroughly analyze, document, and publish the results of our system performance during the examination of the subject CASS specimens.

Figure 5. Pressurizer and Surge Line Details

CASS PIPING SYSTEM APPLICATIONS
Typical CASS Piping Weld Locations in PWR Reactor Coolant Systems
The following graphic illustrates the location and extent of CASS materials in the RCS of many PWR plants.

RCS Main Loop Piping Welds: This portion of the RCS contains large diameter butt welds that join centrifugally cast stainless steel (CCSS) piping segments to statically cast stainless steel (SCSS) elbows and reactor coolant pump (RCP) casings. RCS main loop piping includes the following subassemblies:

  • Hot leg piping from the Reactor Vessel Outlet to the SG Inlet
  • Cross-over piping from the SG Outlet to the RCP Inlet
  • Cold leg piping from the RCP Outlet to the RPV Inlet

Steam Generator Inlet / Outlet Nozzle DM Welds: These terminal end DM butt welds are present in PWR plants, both with and without safe ends between the SCSS elbows and the ferritic steel nozzle forgings. 

Pressurizer Surge Piping Welds: This portion of the RCS contains a series of butt welds fabricated using CCSS piping segments to SCSS elbows between the Pressurizer Surge nozzle end and the Hot Leg Surge nozzle. 

SUMMARY
The CASS piping welds present in many PWR plants provide numerous and complicated challenges to their effective ultrasonic examinations. SI’s new CASS ultrasonic examination system will provide a new and meaningful solution.

PROJECT TIMELINE
SI is working to complete the development, integration and capability demonstrations of the CASS ultrasonic examination system described in this document for limited (emergent) fall 2023 and scheduled deployments beginning in spring 2024.

Get News & Views, Volume 53

Contact Us

News & Views, Volume 53 | PEGASUSTM Nuclear Fuel Code

/in , , , , /by

ADVANCED FUEL MODELING DEVELOPMENT STATUS

By:  Bill Lyon

INTRODUCTION
The PEGASUS nuclear fuel behavior code features a robust 3D, finite element modeling (FEM) computational foundation capable of performing both thermo-mechanical and structural non-linear analyses within a highly versatile and customizable computational platform. The first applications of PEGASUS were for light water reactor (LWR) fuels and materials. Development work on PEGASUS has been extended to advanced fuel designs such as those proposed for Advanced Technology Fuel (ATF) LWR applications and Gen IV reactor designs, including gas and liquid metal-cooled reactors (GCRs and LMRs). 

The versatility and adaptability of PEGASUS is key in enabling extensions to non-conventional operating environments, materials, fuel forms, and geometries.

LWR APPLICATIONS
SiC Cladding

Figure 1. SiGA cladding is a multi-layered composite design composed of SiC fiber in a SiC matrix.

A project is underway to further the development and irradiation testing of a composite silicon carbide matrix as an ATF cladding material. This research is supported through a DOE Funding Opportunity award (DE-FOA-0002308) for the irradiation of a composite silicon-carbide (SiC) ceramic matrix material in an existing U.S. commercial LWR. This work is led by General Atomics – Electromagnetic Systems (GA-EMS) with Structural Integrity Associates (SIA) as a primary subcontractor. For this work, PEGASUS is being adapted to model monolithic and composite SiC manufactured by GA-EMS, SiGA [1], through the incorporation of proprietary material constitutive models. PEGASUS will then be used to provide independent test performance analyses aiding in the design of the irradiation vehicle and predicted material performance. The goal of the testing is to gather irradiation data under prototypic LWR operating conditions and to inform and confirm material performance models for the SiGA-based cladding. A follow-on activity is planned to evaluate the predicted performance compared to data gathered during the post-irradiation examination phase of the project.

Figure 2. Lightbridge Fuel Design PEGASUS Models

Cruciform Metallic Fuel
An additional fuel concept that has been explored using PEGASUS is a cruciform, extruded metallic fuel design proposed by Lightbridge Corporation [2]. This fuel is characterized by a unique multi-lobed fuel cross-section and features a U-50Zr fuel composition. Recent work has been published on fabrication testing of this proposed fuel design by Pacific Northwest National Laboratory (PNNL) [3]. PEGASUS has been used previously to prototype 2D and 3D geometric models and meshes of Lightbridge fuel and to perform fundamental temperature and stress distributions for this fuel under prototypic LWR conditions. PEGASUS has specific modeling tools designed to facilitate “extruded” 3D fuel designs that automate the meshing of these geometries. More work in this area is planned as a proposal has recently been awarded under the DOE NEUP program (DE-FOA-0002732) funding a collaborative project led by Texas A&M University along with Lightbridge, NuScale, and Structural Integrity Associates, Inc. (SI) for modeling this type of fuel for application in a LWR SMR.

 

URANIUM METAL ALLOY FUELS FOR SODIUM-COOLED FAST REACTORS
The initial implementation of metallic alloy fuel and stainless-steel cladding material constitutive models for prototypic fast reactor fuel designs in PEGASUS has been completed. Material properties and behavioral models for U-Pu-Zr fuel and HT-9 (high Chromium, martensitic stainless steel) cladding have been added. Ongoing work includes the implementation of a gaseous swelling and fission gas release behavior model for U-Pu-Zr fuel, a Zr-redistribution model, and a fuel-cladding chemical interaction (FCCI) model that includes the effect on cladding wall thinning.

To test the implementation of these models, benchmark tests were prepared that provided comparative data for assessment of the models’ performance. Test cases were chosen from two experimental series irradiated in EBR-II: X430, a 37-pin hexagonal sub-assembly, and X441, a 61-pin bundle. These experiments were designed to test numerous fuel rod design variables and fuel response as a function of fuel alloy composition, smear density, plenum-to-fuel volume ratio, power, and coolant conditions [4]. The general experimental fuel rod design corresponds to the typical driver fuel configuration shown in Figure 3.

Figure 3. Typical EBR-II Mark-III or Mark-IIIA Fuel Element [5]

Figure 4. Left: 2D Computational Model of Rod DP2, Right: Temperature Contour Plot of the Fuel Stack Region for Rod DP21 at Peak Power (plenum region removed for detail)

An illustration of the model and selected results from the initial analysis of rod DP21, assembly X441 are shown in the figures above. Figure 4 provides a diagram of the computational model showing the primary components of the model and a plot of the temperature distribution throughout the fueled region of the rod at peak power. Figure 5 provides the radial temperature profile across the fuel rod from the center to the cladding outer surface at peak power near the end of the irradiation period. Temperatures vary from just ~900 K at the pellet center to ~650 K at the cladding surface. The temperature differential is fairly low at ~250 K, as would be expected from a high-conductivity metal fuel rod with a Na-bonded fuel cladding gap. These results are consistent with published experimental observations.

TRISO FUEL MODELING DEVELOPMENT
Several advanced fuel material models have been implemented specifically for TRISO fuel in PEGASUS, including thermal and mechanical models for UCO or UO2 kernels, PyC, SiC materials, and a fission gas release model for computing the release of gaseous fission products such as Xe and Kr. In addition to the standard 3D and 2D axisymmetric modeling FEM capabilities in the code, PEGASUS contains several unique tools designed specifically to support TRISO fuel modeling and analysis. These include a “spherical mesh object” tool that can automate the process of generating 2D/3D TRISO spheres, meshing them, and embedding them into a fuel matrix to allow modeling of individual TRISO kernels or fully encapsulated TRISO fuel forms. An example of models generated using the spherical mesh object tool is shown in Figure 6. The spherical mesh object capability is, to our knowledge, unique to PEGASUS and not found in any other fuel performance or general-purpose FEM code. PEGASUS also has a “reshape” function that can automate the process of meshing and modeling deformed TRISO particles to increase user efficiency. Figure 7 illustrates particle meshes that were created using the reshape meshing tool.

These modeling capabilities allow PEGASUS to be used to investigate very detailed mechanical and structural effects in TRISO fuel forms. For example, enabling the detailed analysis of the mechanical interaction between TRISO fuel layers explicitly examining the effects of cracking, debonding, and asphericity within whole or damaged particles.

Figure 5. Radial Temperature Distribution Across the Fuel Rod Model at ~ 486 Days of Irradiation

Planned future development work includes the integration of damage-mechanics modeling and fission product diffusion in the TRISO particle, fuel compact, ad matrix. One failure mode of particular interest that has been identified is cracking of the IPyC layer which propagates through the SiC outer layer. This can create a pathway for enhanced fission product release from the TRISO particle to the surrounding fuel matrix. This failure mechansim appears to occur when the buffer layer remains bonded to the IPyC layer providing the conditions for a synergistic mechanical and chemical failure mechanism that combines cracking, stress concentration, and chemical corrosion (localized Pd-induced corrosion in the SiC [6]. This failure mode is of interest because it can have a strong impact on fuel source term determination for operational TRISO fuel.

Figure 6. Temperature distribution in a cross-section of a 3D slab of a TRISO compact matrix model with a “sparse”, random kernel distribution under prototypic gas-cooled reactor conditions. (Generated using the “spherical mesh object” tool.)

SUMMARY
PEGASUS is an advanced analysis tool developed for industry applications that can provide a complimentary and independent capability for nuclear fuel performance. Recent development work on PEGASUS has focused on expanding the applicability of the code to the advanced fuel (ATF) and advanced reactor arena. Future development is planned for PEGASUS that will continue along multiple avenues with an emphasis on advanced fuels and specific thermo-mechanical issues within the industry, such as deterministic failure model development. One example of this is the aforementioned Pd-induced failure mechanism identified for TRISO fuel. SI is actively seeking partners within the advanced fuel community to collaborate with on this work and would welcome inquiries and proposals for expanded application of PEGASUS.

 

Figure 7. Deformed 3D TRISO particle meshes generated using the “reshape” function tool in PEGASUS.

References

  1.   C. P. Deck et al., Overview of General Atomics SiGA™ SiC-SiC Composite Development for Accident Tolerant Fuel, Transactions of the American Nuclear Society, Vol. 120, Minneapolis, Minnesota, June 9–13, 2019.
  2.   J. Malone, A. Totemeier, N. Shapiro, and S.  Vaidyanathan, Lightbridge Corporation’s Advanced Metallic Fuel, Nuclear Technology, Vol 180, Dec. 2012.
  3.  Z. Huber and E. Conte, Casting and Characterization of U-50Zr, PNNL-33873, Pacific Northwest National Laboratory, Richland, Washington 99354, Jan. 2023.
  4.  C. E. Lahm, J. F. Koenig, R. G. Pahl, D. L. Porter, and D. C. Crawford, “Experience with Advanced Driver Fuels in EBR-II,” J. Nucl. Mat. 204 (1993) 119-123.
  5. G. L. Hofman, M. C. Billone, J. F. Koenig, J. M. Kramer, J. D. B. Lambert, L. Leibowitz, Y. Orechwa, D. R. Pedersen, D. L. Porter, H. Tsai, and A. E. Wright. Metallic fuels handbook, Technical Report ANL-NSE-3, Argonne National Laboratory, 2019.
  6. J.D Hunn, C.A. Baldwin, T.J. Gerzak, F.C. Montgomery, R.N. Morris, C.M. Silva, P.A Demkowicz, J.M. Harp, and S.A. Ploger, Detection and Analysis of Particles with Failed SiC in AGR-1 Fuel Compacts, Nuclear Engineering and Design, April 2016.

Get News & Views, Volume 53

Contact Us
News & Views, Volume 49 | Inspection Optimization- Probabilistic Fracture Mechanics

News & Views, Volume 49 | Inspection Optimization: Probabilistic Fracture Mechanics

/in , , , , , , , /by

By:  Scott Chesworth (SI) and Bob Grizzi (EPRI)

News & Views, Volume 49 | Inspection Optimization- Probabilistic Fracture Mechanics

The goal was to determine whether the frequency of current inspection requirements was justified or could be optimized (i.e., increase the interval of certain inspections to devote more attention to higher-value inspections and thereby maximize overall plant safety).

Executive Summary
Welds and similar components in nuclear power plants are subjected to periodic examination under ASME Code, Section XI.  Typically, examinations are performed during every ten-year inspection interval using volumetric examination techniques, or a combination of volumetric and surface examination techniques.  Nuclear plants worldwide have performed numerous such inspections over plant history with few service induced flaws identified.  

SI was selected by the Electric Power Research Institute (EPRI) to review the technical bases for the inspection intervals for select components.  The goal was to determine whether the frequency of current inspection requirements was justified or could be optimized (i.e., increase the interval of certain inspections to devote more attention to higher-value inspections and thereby maximize overall plant safety.)  

An inspection interval review takes into consideration industry operating experience (OE), operating history and previous inspection data.  Many of the components / welds are difficult to access (require scaffolding and removing insulation), require manual techniques of inspection, and are typically in high radiological dose areas.  The inspections can also have significant impact to outage duration.  Reducing the frequency of inspections has the potential for time and cost savings during outages and reduces the radiation exposure to plant personnel.  From the inspection interval review, one utility noted that increasing the inspection interval for steam generator nozzle welds from 10 years to 30 years would save over $600,000 of inspection and supporting activity costs over a 60-year licensed period of operation.  Actual savings for a given plant are situation-dependent, although the potential for significant Operations and Maintenance (O&M) savings exists.

Background

To identify which components and inspection requirements were most suitable for optimization, EPRI performed an initial scoping investigation to collect the following information:

  • The original bases for the examinations, if any;
  • Applicable degradation mechanisms, and the potential to mitigate any potential damage associated with each mechanism;
  • Operating experience, examination data, and examination results, e.g., fleet experience;
  • Previous relief requests submitted to regulators;
  • Industry guidance documents that replace or complement ASME Code requirements;
  • Redundancy of inspections caused by other industry materials initiatives and activities (e.g., Boiling Water Reactor Vessel and Internals Project (BWRVIP), Materials Reliability Program (MRP), etc.); and
  • Existing ASME Code Cases that provide alternatives to existing ASME Code inspection requirements and their bases.

After compilation and review of the information collected, EPRI and their members determined that the inspection requirements for the following components were among the most suitable for optimization:

  • Pressurized water reactor (PWR) steam generator shell and nozzle welds and nozzle inside radius sections;
  • PWR pressurizer shell and nozzle welds; and
  • Boiling water reactor (BWR) heat exchanger shell and nozzle welds and nozzle inside radius sections.

Once the components were identified, EPRI contracted with SI to support development of the technical bases to optimize the related inspections.  These evaluations are documented in the following four EPRI reports, all of which are publicly available for download at www.epri.com:

  • Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections, EPRI, Palo Alto, CA: 2019. 3002014590.
  • Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 2 Vessel Head, Shell, Tubesheet-to-Head, and Tubesheet-to-Shell Welds, EPRI, Palo Alto, CA: 2019. 3002015906.
  • Technical Bases for Inspection Requirements for PWR Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, EPRI, Palo Alto, CA: 2019. 3002015905.
  • Technical Bases for Examination Requirements for Class 2 BWR Heat Exchanger Nozzle-to-Shell Welds; Nozzle Inside Radius Sections; and Vessel Head, Shell, and Tubesheet-to-Shell Welds, EPRI, Palo Alto, CA: 2020. 3002018473.

Why It Matters

Recent efforts in the nuclear industry include a focus on reducing the cost of generating electricity to make nuclear more competitive with other sources (natural gas, etc.).  A major component of these efforts is a targeted reduction of plant O&M costs, while ensuring that there is no detrimental impact on plant safety.  Reducing low-value (i.e., low-risk, high-cost) inspections allows plant resources to be devoted to higher value activities (e.g., preventative maintenance).  This is one benefit of employing risk-informed approaches.

The industry (in conjunction with EPRI, SI, and others) has shown a great deal of interest in employing risk-informed approaches where appropriate.  Such efforts include (but are not limited to):

Extremely Low Probability of Rupture (xLPR)

  • ASME Code Case N-702 (alternative requirements for BWR nozzle inner radius and nozzle-to-shell welds)
  • ASME Code Case N-711 (volume of primary interest)
  • ASME Code Case N-716-1 (streamlined risk-informed inservice inspection)
  • ASME Code Case N-752 (risk-informed repair / replacement)
  • ASME Code Case N-770-6 (cold leg piping dissimilar metal butt weld inspection)
  • ASME Code Case N-864 (reactor vessel threads in flange examinations)
  • ASME Code Case N-885 (alternative requirements for interior of reactor vessel, welded core support structures and interior attachments to reactor vessels, and removable core support structures)
  • ASME Code Case N-[xxx] (alternative requirements for pressure-retaining bolting greater than 2 inches in diameter)
  • 10CFR50.69 (risk-informed categorization and treatment of systems, structures, and components)
  • The inspection optimization approach discussed here is congruent with these other approaches, as it uses probabilistic and risk insights to help plants to prioritize inspection and maintenance activities on those components most significant to plant safety.

How It’s Done

In the four EPRI reports cited above, the technical basis for increasing the interval of components inspections included the following steps:

  • Review of previous related projects
  • Review of inspection history and examination effectiveness
  • Survey of components and selection of representative components for analysis
  • Evaluation of potential degradation mechanisms
  • Component stress analysis

Once the above steps were completed, components are subjected to Deterministic and Probabilistic Fracture Mechanics Evaluations.  The DFM and PFM approaches used in the EPRI reports are based on methods used in previous inspection optimization projects, and involved either an increase in examination interval, a reduction in examination scope, or both.  The DFM evaluations were performed using bounding inputs to determine the length of acceptable component operability with a postulated flaw.  The results of the DFM investigation were also to determine the critical stress paths for consideration in the PFM analyses.  The results of the DFM evaluations concluded that all selected components are very flaw tolerant, with the capability of operating with a postulated flaw for more than 80 years.

PFM evaluations were performed to demonstrate the reliability of each selected component assuming various inspection scenarios (e.g., preservice inspection (PSI) only, PSI followed by 10-year in-service inspections (ISI), etc.).  Monte Carlo probabilistic analysis techniques were used to determine the effect of randomized inputs and various inspection scenarios on the probabilities of rupture and leakage for the selected components.  Sensitivity studies are performed to investigate possible variation in the various input parameters to establish the key parameters that most influence the results. 

For each component, probabilities of rupture and leakage were determined for the limiting stress paths in each selected component for a variety of inspection scenarios.  The results of the PFM evaluations demonstrated that the NRC acceptance criteria of 1.0E-6 for both probabilities of rupture and leakage could be maintained for all components for inspection intervals longer than the 10-year intervals defined in Section XI of the ASME Code.  Therefore, the results demonstrate that examinations for the selected components can be extended beyond current the ASME Code-defined interval; in some cases, they can be extended out to the end of the current licensed operating period (at least 30 years for most plants).

Why Structural Integrity

SI is the primary author of the four EPRI Reports cited above (3002014590, 3002015906, 3002015905 and 3002018473).  The inspection optimization projects have provided SI with the opportunity to use its experience in structural reliability to develop a customized PFM software tool named PROMISE (PRobablistic OptiMization of InSpEction), which was used to optimize the inspection schedules for various plant components.  The PROMISE software implemented a probabilistic model of fatigue crack growth using linear elastic fracture mechanics (LEFM) methods, consistent with ASME code, Section XI flaw evaluation procedures.  

The software was developed, verified & validated (V&V), and tested under the provisions of a 10 CFR 50, Appendix B Nuclear Quality Assurance Program.  This tool is based on other, similar previous software codes, and it can be used for similar applications in the nuclear industry where a rigorous technical basis is required to optimize inspection schedules for high-reliability components involving significant outage impact.  In 2020, the NRC staff conducted an audit of PROMISE.  According to the conclusion of the audit report (ML20258A002), the NRC staff gained a better understanding of how PFM principles were implemented in PROMISE and of the V&V on the software.  

In addition to the software audit, SI has supported EPRI and industry in developing responses to NRC requests for additional information (RAIs) for the pilot plant submittals for all four EPRI Reports.  This experience has given SI a great deal of understanding regarding the most efficient and effective way to preemptively address potential NRC concerns in future plant-specific submittals.

How It Would Work For You

For plant owners to use the technical bases established by this work to obtain relief for their plant, they must demonstrate that the representative geometries, materials, and loading conditions used for the selected components bound their plant-specific information.  Based on this analysis, the EPRI Reports provide criteria for each component regarding the component configuration, component dimensions, component materials, applicable transient loadings, and other relevant parameters that must be satisfied on a plant-specific basis.  If all criteria are satisfied on a plant-specific basis for a given component, the results of the investigation can be used for the plant as the technical basis to establish revised inspection schedules for that component.  If any criteria are not satisfied, then plant-specific analysis is required to address any unbounded conditions.  SI can provide support in several areas, including:

Since the technical basis in the EPRI Reports used generic plant configurations, some plant configurations were not included in the analysis.  SI can also support efforts by plants with such configurations to determine whether they are bounded by the criteria of the EPRI Reports.

  • Evaluation of plant-specific parameters against report criteria to determine whether a given plant configuration is bounded
  • Performing plant-specific analysis (e.g., component stress analysis, DFM and PFM, etc.) required to address any unbounded conditions
  • Supporting development of the relief request to proactively address known NRC areas of concern
  • Supporting development of responses to any NRC requests for additional information

Plant Experience To Date

The first plant-specific submittal was made by a U.S. two-unit PWR site in December 2019 based on EPRI Report 3002014590, requesting an inspection alternative to current ASME Code, Section XI examination requirements for steam generator main steam and feedwater nozzle-to-shell weld and inner radii examinations.  The alternative requests an increase in the inspection interval for these items from 10 to 30 years.  The safety evaluation report (SER) for this alternative was received from the NRC in January 2021.

The first plant-specific submittal was made by a U.S. PWR site in December 2019 based on EPRI Report 3002015906, requesting an inspection alternative to current ASME Code, Section XI examination requirements for steam generator Class 1 nozzle-to-vessel welds and Class 2 vessel head, shell, tubesheet-to-head, and tubesheet-to-shell welds.  The alternative requests an increase in the inspection interval for these items from 10 to 30 years.  RAIs for this alternative were received from the NRC in February 2021.  SI supported development of the RAI responses.

The first plant-specific submittal was made by a U.S. two-unit PWR site in December 2019 based on EPRI Report 3002015905, requesting an inspection alternative to current ASME Code, Section XI examination requirements for Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds.  The alternative requests an increase in the inspection frequency for these items from 10 to 30 years.  RAIs for this alternative were received from the NRC in February 2021.  SI supported development of the RAI responses.

The first plant-specific submittal was made by a U.S. two-unit BWR site in December 2019 based on EPRI Report 3002018473, requesting an inspection alternative to current ASME Code, Section XI examination requirements for Class 2 BWR heat exchanger nozzle-to-shell welds; nozzle inside radius sections; and vessel head, shell, and tubesheet-to-shell welds.  The alternative requests an increase in the inspection interval for these items from 10 years to the end of the plant’s current operating license.  RAIs for this alternative were received from the NRC in February 2021.  SI supported development of the RAI responses.

Conclusion

Inspection optimization offers the opportunity to reallocate plant resources to higher value activities.  In a highly competitive electricity market, the work here has shown opportunity exists to improve O&M costs and maintain safety through effective analysis.  

SI brings to bear the prior experience in developing the methodology with EPRI, proprietary NQA-1 verified software, and decades of industry credibility to support all aspects of the efforts required to institute a program of inspection optimization.  

Get News & Views, Volume 49

Contact Us
News & Views, Volume 49 | Autobook- Nuclear Physics Automation Code

News & Views, Volume 49 | Autobook: Nuclear Physics Automation Code

/in , , , , /by

News & Views, Volume 49 | Autobook- Nuclear Physics Automation CodeBy:  Sasan Etemadi, P.E. and Mark Drucker, P.E.

The AUTOBOOK code reduces human errors, increases efficiency, and streamlines the reload analysis process

AUTOBOOK facilitates plant operation by providing nuclear power plant Reactor Engineers and Reactor Operators with cycle-specific information about the physics characteristics of the reactor core in a core data book document. Structural Integrity has created the AUTOBOOK computer code to automate the creation of this document.

AUTOBOOK is a Quality Assured code developed under a licensee’s software quality assurance (SQA) program. SI provides a full complement of SQA documents, including a Software Requirement Specification (SRS), a Software Design Description (SDD), Verification and Validation (V&V) Plan and Test Report, a User Manual, and Software Installation Instructions (SII).

READ MORE

News & View, Volume 48 | Increase in Reinspection Intervals for BWR Reactor Internals

News & Views, Volume 48 | Increase in Reinspection Intervals for BWR Reactor Internals

/in , , , , /by

By:  Dick Mattson and Minghao QinNews & View, Volume 48 | Increase in Reinspection Intervals for BWR Reactor Internals

A U.S. BWR utility contracted with Structural Integrity (SI) to review their current reinspection guidance documents relative to those contained in the BWRVIP inspection guidelines, the purpose of which was two-fold:

  1. ­Are current reinspection guidelines compliant with industry requirements?
  2. ­Are there components where reinspection intervals could possibly be extended?

READ MORE

News & View, Volume 48 | Fatigue Adjustment Factors for Increased Cyclic Life

News & Views, Volume 48 | Fatigue Adjustment Factors for Increased Cyclic Life

/in , , , , /by

By:  Bill WeitzeNews & View, Volume 48 | Fatigue Adjustment Factors for Increased Cyclic Life

100% of thermal stress was treated as nonlinear gradient stress and linear bending stress was about 12% of the moment stress. Structural Integrity’s (SI’s) review of the stress terms used in piping analysis show that pressure stress does create bending stress in components…

EPRI Report 3002014121 “Development of Fatigue Usage Life and Gradient Factors” has developed fatigue usage adjustment factors that account for: 1) increased cyclic life associated with the growth of potential engineering size fatigue cracks in thicker components (thickness factor, TF; also called life factor, LF), and 2) the presence of through-thickness stress gradients (gradient factor, GF). (TF is used in the issued Code Case.)  These factors are applied to cumulative usage factor, U, in air.

READ MORE

News & View, Volume 48 | Examination Optimization for PWR and BWR Components

News & Views, Volume 48 | Examination Optimization for PWR and BWR Components

/in , , , , , /by

By:  Scott Chesworth, Bob Grizzi, and Dilip Dedhia

Optimizing the inspection interval for high-reliability components whose examinations have a significant outage impact.News & View, Volume 48 | Examination Optimization for PWR and BWR Components

Welds and similar components in nuclear power plants are subject to periodic examination under ASME Code, Section XI.  Typically, examinations are performed during every ten-year inspection interval using volumetric examination techniques, or a combination of volumetric and surface examination techniques.  Nuclear plants worldwide have performed numerous such inspections over the plant history with few service induced flaws identified.  Since personnel health and safety, radiation exposure, and overall outage costs associated with these inspections can be significant, Structural Integrity (SI) was contracted by the Electric Power Research Institute (EPRI) to review the technical bases for the inspection intervals for select components.  The goal was to determine whether the frequency of current inspection requirements was justified or could be optimized (i.e., reduced in order to devote more attention to higher-value inspections and thereby maximize overall plant safety).  Special priority was given to components demonstrating an exceptional history of reliability and whose examinations have a significant outage impact.

News & Views, Volume 48 | Environmentally-Assisted Fatigue Screening and Managing EAF Effects in Class 1 Reactor Coolant Components

News & Views, Volume 48 | Environmentally-Assisted Fatigue – Screening and Managing EAF Effects in Class 1 Reactor Coolant Components

/in , , , , /by

By: Dave Gerber and Terry HerrmannNews & Views, Volume 48 | Environmentally-Assisted Fatigue Screening and Managing EAF Effects in Class 1 Reactor Coolant Components

Environmentally-Assisted Fatigue (EAF) screening is used to systematically identify limiting locations for managing EAF effects on Class 1 reactor coolant pressure boundary components wetted by primary coolant.  This article provides an overview of the methods developed and used by Structural Integrity (SI) for Class 1 components having explicit fatigue analyses performed using ANSI/ASME B31.7(1) and ASME Section III(2).  A future article will discuss how this is performed for Class 1 piping designed and analyzed to ASME/ANSI B31.1(3).

READ MORE

News & View, Volume 48 | SI-FatiguePro Version 4.0 Crack Growth Module Application Case Study; Complex Multi-Cycle Nuclear Transients

News & Views, Volume 48 | SI:FatiguePro Version 4.0 Crack Growth Module – Application Case Study Complex Multi-Cycle Nuclear Transients

/in , , , , /by

News & View, Volume 48 | SI-FatiguePro Version 4.0 Crack Growth Module Application Case Study; Complex Multi-Cycle Nuclear Transients

By: Curt Carney

As plants enter their initial or subsequent license renewal period one of the requirements is to show that fatigue (including environmental effects) is adequately managed.  For some locations in pressurized water reactors (PWRs), it can be difficult to demonstrate an environmental fatigue usage factor less than the code allowable value of 1.0.  Therefore, plants are increasingly turning to flaw tolerance evaluations using the rules of the ASME Code, Section XI, Appendix L.  Appendix L analytically determines an inspection interval based on the time it takes for a postulated flaw (axial or circumferential) to grow to the allowable flaw size.  For surge line locations, this evaluation can be very complex, as the crack growth assessment must consider many loadings, such as: insurge/outsurge effects, thermal stratification in the horizontal section of the line, thermal expansion of the piping (including anchor movements), and internal pressure.  Trying to envelope all of these loads using traditional tools can lead to excess conservatism in the evaluation, and short inspection intervals that reduce the value of an Appendix L evaluation.

News & View, Volume 47 | TRU Compliance Expands into Physical Security | How To Make Knowing A Good Thing - Thinning Handbooks

News & View, Volume 47 | How To Make Knowing A Good Thing: Thinning Handbooks

/in , , , /by

By:  Stephen Parker and Eric HoustonNews & View, Volume 47 | TRU Compliance Expands into Physical Security | How To Make Knowing A Good Thing - Thinning Handbooks

SI has developed a process to mitigate the negative outcomes of piping examination.  One part of that process is Thinning Handbooks, which have resulted in direct savings in excess of $10 Million for one nuclear plant.

Examination of Safety Related Service Water piping is driven by a number of factors, all of which tend to converge on the objective of finding localized thinning prior to the thinning becoming a problem.  In other words, examinations are performed to eliminate the risk of a leak and ensure that the wall thickness remains greater than tmin (the minimum required uniform wall thickness).  However, the rules, regulations, and economic realities mean that only bad things happen from an exam regardless of what is found.

READ MORE