Robert Carter

Structural Capability of Austenitic Piping and Shells With Parallel Offset Cracks

Boiling water reactor (BWR) components made of austenitic stainless steel are exposed to the high temperature water environment. Under the combination of susceptible material condition (sensitization due to welding or cold work), weld residual stresses and the oxygenated high temperature water environment, reactor internals such as the core shroud and internal core spray piping experience stress corrosion cracking. The cracking is typically in the weld heat affected zone on both sides of the weld and is parallel to the weld direction. The determination of the structural capability of the component with parallel offset cracks is the subject of this paper. The Section XI, ASME code position on parallel planar flaws states that there is no interaction (i.e. no change in load capability due to multiple parallel cracks) for cracks that are separated by 12.5 mm (1/2 inch) or more. While this is reasonable under linear elastic fracture mechanics conditions, it is not conservative for ductile failure under limit load conditions. Alternatively, assuming that the parallel cracks are in a single plane for the purpose of determining the load capability is over-conservative and underestimates the structural capability. This paper is an extension of earlier work on the interaction of parallel flaws and considers the combines load capability as a function the crack separation. The interaction rules developed here were based on analysis and validated by comparison with extensive test data from different sources. For cracks in two parallel planes, the proposed rules for ductile materials (where limit load is the governing failure mechanism) allow them to be considered as separate cracks if the distance between the two planes is greater than 3t (where t is the shell thickness) and the limit load for cracking in each plane is determined separately. For planes separated by less than t, they are combined and assumed to be in one plane and the limit load is calculated for the combined crack. Linear interpolation is used when the separation distance, d is such that t≤d ≤3t.Copyright

Technical Basis for BWRVIP Stainless Steel Crack Growth Correlations in BWRs

An intergranular stress corrosion cracking (IGSCC) growth model for unirradiated, thermally sensitized stainless steels and a methodology for assessment of through-wall crack growth in horizontal weld heat affected zones (HAZs) of boiling water reactor (BWR) core shrouds has been developed. This empirical model accounts for the variability of important IGSCC parameters such as coolant conductivity, stress intensity factor, K, temperature and electrochemical corrosion potential (ECP) in providing a conservative, yet realistic assessment of the crack growth rate. Data from various sources were used to derive the empirical crack growth correlation, including work from Electric Power Research Institute (EPRI)-sponsored research, work sponsored by the U. S. Nuclear Regulatory Commission (NRC) and in-plant crack arrest verification system (CAVS) data as well as laboratory data developed by the General Electric Nuclear Energy (GENE). The combined database from all the sources was evaluated to ensure that only relevant data was used in the model development. This refined database was used to derive the crack growth correlation using pattern recognition and multivariate modeling tools. For practical application to crack growth evaluation of stainless steel components, three approaches were developed for dispositions of flaws, i.e., a K-independent approach, a conservative 95th percentile K-dependent approach and a plant specific approach using actual BWR water chemistry data. An example problem representing actual BWR shroud conditions is presented that demonstrates how the current crack growth model, the weld residual stress and K can be used to perform a plant-specific evaluation of flawed shrouds. The results of this example demonstrated that significant operating periods are likely for most flawed conditions in BWR core shroud welds before ASME Code Section XI core shroud safety margins are challenged.Copyright

SCC and Fracture Toughness of XM-19

The effect of stress intensity factor, cold work, corrosion potential and water purity on the stress corrosion crack (SCC) growth rate behavior of as-received and as-received plus 19.3% cold worked XM-19 was investigated in 288 °C BWR water. For 19.3% cold rolled XM-19, high to very high crack growth rates were consistently observed at high corrosion potential, largely independent of heat or orientation. As received XM-19 exhibited SCC growth rates ~5–10X slower than cold worked XM-19, but these rates are still considered high. For all materials and conditions, low corrosion potential conditions reduced the growth rates by about an order of magnitude, and somewhat more if impurities were present in the water. The SCC growth rates for both conditions of XM-19 were somewhat higher than the equivalent conditions of 18-8 stainless steels, such as Types 304/304L and 316/316L. Higher growth rates tend to be observed at higher yield strength, and XM-19 has an elevated yield strength from nitrogen-strengthening; incomplete annealing in the as-received material can also increase the yield strength. The J-R fracture resistance of 19.3% cold rolled XM-19 and as-received XM-19 in multiple orientations and with replicates was evaluated in 288 °C air. The data show a significant effect of crack orientation in the plate (the rolling plane coincides with the plane of the plate), consistent with the inhomogeneous nature of the microstructure. The fracture resistance of as-received XM-19 was good, but the 19.3% cold rolled XM-19 specimens exhibited low toughness, to the extent that many tests were invalid. Fracture resistance in 80–288 °C water environments was not evaluated, but is relevant to LWR components. Irradiation of this heat of XM-19 is in progress at the Idaho National Laboratory Advanced Test Reactor.

Thermal Aging and Neutron Embrittlement Evaluation of Cast Austenitic Stainless Steels

NUREG-1801, Rev. 1, Section XI.M.13, states that an ASME Code Section XI VT-3 examination is required to be performed of reactor internal components. In addition, the NUREG report specifies that for the license renewal period, these inspections shall be augmented by an aging management program to assess the synergistic effects of thermal aging and neutron embittlement of cast austenitic stainless steel (CASS) components. This aging management program consists of (a) identifying susceptible components; and (b) either performing additional inspections of these components, or performing a component-specific evaluation to confirm that the stresses in the components are sufficiently low such that augmented inspections are not warranted. This paper presents the results of evaluations conducted to assess the potential synergistic effects of thermal aging and neutron embrittlement of CASS components in BWR internals and recommend augmented inspections if needed. The evaluation shows that all the BWR CASS components have ferrite levels below the level for which aging embrittlement is a concern. Furthermore, for the Control Rod Guide Tube Base and Core Spray Sparger Nozzle Elbows, the end of life fluence is less than the threshold value for toughness loss. The end-of-life fluence levels for the orificed fuel support, the jet pump assembly castings and the Low Pressure Core Injecion (LPCI) Couplings exceed the threshold, but the toughness data for irradiated austenitic stainless steel show that these components will have sufficient fracture toughness at end of the license renewal period so that augmented inspection is not required. It is concluded that augmented inspections are not required for the BWR CASS internals.Copyright

In-Service Inspection Strategy for Alloy X-750 BWR Jet Pump Beams Based Upon Linear Elastic Fracture Mechanics Analysis

Jet pumps in a boiling water reactor (BWR) are located in the annulus region between the core shroud and the reactor vessel wall and provide core flow to control reactor power. Between 16 and 24 jet pumps are included in BWR/3 through BWR/6 plants, depending on the plant rating. The inlet mixer assembly of the jet pump is secured in place with a hold down mechanism called a jet pump beam. This beam is fabricated of alloy X-750 and tensioned to 58–74% of the yield stress of the material, depending on the beam design. In recent years, more attention has been placed upon inter-granular stress corrosion cracking (IGSCC) of alloy X-750 BWR internal components as a result of in-service cracking and failures. BWR plant owners have implemented actions to manage IGSCC of jet pump beams and assemblies through increased inspections and changes to process specifications for X-750. However, a thorough understanding of the flaw tolerance of the jet pump beam was not available to guide the periodicity of inspections as well as to define critical flaw sizes needed to validate the capability of inspection techniques. This paper describes a linear elastic fracture mechanics (LEFM) evaluation in which the flaw tolerance of the existing jet pump beam designs is established and used to recommend inspection frequencies for the jet pump beam. Industry operating experience is used to assess the credibility of the results obtained from this evaluation. This work illustrates an example of the use of LEFM to develop a technically defensible basis for the required inspection regions and the frequency of inspection for an alloy X-750 BWR internal component and helps to establish the necessary sensitivity of non-destructive examination technology to be used to examine the component.Copyright

Nickel Alloy Crack Growth Correlations in BWR Environment and Application to Core Support Structure Welds Evaluation

An intergranular stress corrosion cracking (IGSCC) growth model for unirradiated nickel-base alloys (Alloys 82, 182 and 600) in boiling water reactor (BWR) environments has been developed by EPRI. This model has been used for assessment of the crack growth rates in BWR nickel base austenitic alloys with particular application to the BWR shroud support structure materials and welds, including attachments to the reactor pressure vessel fabricated from these alloys. However, the crack growth model can be used for other components with like materials in BWR environments provided that specific parameters such as stresses and stress intensity factor (KI ) distributions are determined. The methodology involves development of crack growth disposition curves that can account for the variability of important IGSCC parameters to provide a conservative, yet realistic assessment of crack growth rate in BWR environments. An extensive nickel base alloy crack growth rate database was developed from data generated through the peer review process and includes both experimental data points and in-plant crack arrest verification system data. Most of the data in the database have reasonable definition of environmental conditions and other important crack growth parameters thus permitting a more realistic generic crack growth model to be developed. Although most of the data is for Alloy 182, it bounds the crack growth rate of Alloy 82 and Alloy 600. The database was used to derive crack growth disposition curves under normal water chemistry (NWC) and hydrogen water chemistry (HWC) conditions. The disposition curves have two stress intensity regimes; one for KI 25 ksi√in where the crack growth is KI -independent. The crack growth disposition curves were used together with a crack growth estimation methodology to determine the crack propagation of the BWR shroud support structure welds which are fabricated from Alloy 82/182. The steps involved in the development of the methodology include determination of residual stresses and operating stresses, development of stress intensity factor (KI ) solutions for crack propagation in the through-thickness direction and estimation of crack growth rates. This methodology was applied specifically for crack growth in the through-thickness direction. Application of this crack growth model to BWR shroud support structure welds H8 and H9 indicates that there is an adequate time period between inspections before initial cracks of ≤10% through-wall thickness reaches the allowable flaw sizes, particularly for HWC conditions.Copyright

The Feasibility of Using a Risk-informed Approach for Calculating Reactor Pressure Vessel Heatup and Cooldown Operating Curves

The current procedures for calculating pressure-temperature (P/T) limits for normal reactor startup and shutdown are defined by deterministic fracture mechanics methodology in the ASME Code, Appendix G (in both Section XI and Section III). The recent pressurized thermal shock (PTS) re-evaluation effort used a very thorough probabilistic fracture mechanics (PFM) evaluation to develop a technical basis to increase the PTS screening criteria. The feasibility of applying this same PFM methodology to evaluate normal startup and shutdown operation for both pressurized water reactor (PWR) and boiling water reactor (BWR) pressure vessels is described in this paper. The approach taken in this study was to define a new risk-informed margin term to be applied to the stress intensity factor for membrane tension (KIm) in Appendix G. The margin on KIm was determined by finding the value that results in a vessel failure frequency equal to 10−6 failures/reactor-year when the reactor operates up to the pressure-temperature limits calculated using the risk-informed margin. This simple approach was selected because it results in a minimum change to the current ASME Code procedure, is easy to apply when revising plant operating P/T limits, and allows for an increase in the allowable P/T limits by as much as would be provided by any alternate procedure. The results from this initial study indicate the margin on KIm in Appendix G can be reduced from 2 to 1.5 for PWR vessels for shutdown and potentially reduced from 2 to as low as 1 for startup. The results indicate the margin on KIm in Appendix G can be reduced from 2 to approximately 1 for BWR vessels for shutdown and potentially startup. Additional analyses for PWR and BWR vessels will be needed to develop a comprehensive risk-informed basis for any revisions to the ASME Code, Appendix G.

The BWRVIP Integrated Surveillance Program

Boiling Water Reactor (BWR) plants in the U.S. are designed with radiation surveillance programs. However, the surveillance materials in some plants do not necessarily represent the limiting plate and/or weld material of the reactor pressure vessel (RPV). Also, some plants do not have baseline data for the surveillance materials, which is needed to measure irradiation shift. In 1998 the BWR Vessel and Internals Project (BWRVIP) conceived the BWR Integrated Surveillance Program (ISP) to address these concerns. The ISP surveyed all BWR vessel limiting materials and all available BWR surveillance materials (including materials from a 1990s supplementary research program called the Supplemental Surveillance Program, or SSP). For each vessel limiting weld and limiting plate, a best representative surveillance material was assigned, based on heat number, similar chemistries, common fabricator, and the availability of unirradiated data. Many of the selected surveillance materials are good representatives for the limiting materials of multiple plants, so fewer capsules are required to be tested, reducing the overall cost of surveillance while also improving BWR fleet compliance with 10CFR50 Appendix H.Copyright

EPRI activities to address reactor pressure vessel integrity issues

Abstract The demonstration of reactor pressure vessel (RPV) structural integrity is an essential element in ensuring the continued safe and reliable operation of US nuclear power plants. The Electric Power Research Institute (EPRI), through its domestic and international member utilities, continues to pursue an aggressive research program to develop technologies and capabilities that will address issues associated with reactor pressure vessel integrity. Ongoing research in the EPRI Nuclear Power Group Materials Performance Program covers a broad range of technical areas associated with RPVs. The program is structured under the following product groups; (1) Management and Mitigation; (2) Material Performance Databases; (3) Material Condition Assessment; and (4) Operability Assessment. Specific activities under each of theses product groups are described in this paper.