Flaw Evaluation Procedure for Cast Austenitic Stainless-Steel Materials Using A Newly Developed Statistical Thermal Aging Model

2021 ◽  
Author(s):  
Mohammed F Uddin ◽  
Cédric Sallaberry ◽  
Gery Wilkowski
Keyword(s):  
Statistical Model ◽  
Failure Modes ◽  
Austenitic Stainless Steels ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Chemical Compositions ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Reactor Operating

Abstract Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures leading to very low fracture toughness. Because of their low aged toughness with high variability, flaw evaluations of CASS material need to be established with an understanding of the materials aged condition, especially since most US Pressurized Water Reactor (PWR) nuclear plants have been given plant life extensions for 60-year operation. A flaw evaluation procedure for CASS materials is presented here using a new statistical model developed to predict the toughness of fully aged CASS using the materials' chemical compositions. In this procedure, the Dimensionless-Plastic-Zone-Parameter (DPZP) analysis is used to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) to predict the failure stress for CASS pipe/fittings with a circumferential surface crack. The procedure was validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in US PWR plants -solving 1096 sample problems to understand what range of Z-factors in US PWR plants (for CF8m CASS materials). Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also verified with available test results. This work has been accepted as Code Case N-906 in ASME Boiler and Pressure Vessel (BPV) Code.

Flaw Evaluation Procedure for Cast Austenitic Stainless Steel Materials Using a Newly Developed Statistical Thermal Aging Model

2019 ◽  
Author(s):  
M. Uddin ◽  
C. Sallaberry ◽  
G. Wilkowski
Keyword(s):  
Statistical Model ◽  
Failure Modes ◽  
Austenitic Stainless Steels ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Compositional Variation ◽  
Chemical Compositions ◽  
Test Results ◽  
Thermal Embrittlement ◽  
Reactor Operating

Abstract Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures can lead to a reduction in the fracture toughness and increase in strength. Some aged CASS materials have the potential to have exceedingly low toughness and also show high variability due to the nature of their microstructure or compositional variation within the casting. Because of their low aged toughness with the variability, flaw evaluations of CASS material need to be done with an understanding of the materials aged condition, especially since most US PWR nuclear plants have been given plant life extensions for 60-year operation, and consideration of further extension to 80 years is underway. In this paper, a flaw evaluation procedure for CASS materials is presented using a new statistical model developed to predict the toughness of fully aged CASS using the material’s chemical composition. The new statistical model was developed based on the experimental toughness using standard 1T CT specimens (generally in the L-C orientation) at 288C to 320C and chemical compositions of the CF8m CASS materials. While the detail development of the model is beyond the scope of this paper, a brief validation of predicted toughness using chemical compositions is presented in this paper. Using the predicted toughness, a flaw evaluation procedure was developed using the Dimensionless-Plastic-Zone-Parameter (DPZP) analysis to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) that needs to be applied to the limit-load solution to predict the failure stress for CASS pipe and fittings with a circumferential surface crack. Variability within a single casting was also determined from available test results which was included in the procedure to determine Z-factor. The procedure was then validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes (i.e., limit load or EPFM), etc. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in US PWR plants (whose chemical compositions were known) — essentially solving 1096 sample problems to understand what range of Z-factors might exists in US PWR plants (for CF8m CASS materials) considering all variations in pipe dimensions, ferrite contents, materials’ toughness, etc. Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also investigated by comparing the predictions with available test results.

Comparison of Different Thermal Aging Models to Assess Fully Aged Toughness in Cast Austenitic Stainless Steels

2015 ◽  
Author(s):  
M. F. Uddin ◽  
G. M. Wilkowski ◽  
R. E. Kurth ◽  
F. W. Brust ◽  
D.-J. Shim ◽  
...  
Keyword(s):  
Stainless Steels ◽  
Thermal Aging ◽  
Austenitic Stainless Steels ◽  
Evaluation Procedure ◽  
Chemical Compositions ◽  
Nuclear Plants ◽  
Thermal Embrittlement ◽  
Aging Models ◽  
Reactor Operating ◽  
Relationship Of

Thermal embrittlement of cast austenitic stainless steels (CASS) occurs at reactor operating temperatures during the reactor design lifetime of 40 years leading to a reduction in their toughness and an increase in strength. Additionally most US nuclear plants have been given plant life extensions for 60-year operation, and consideration of further extension to 80 years is underway. As the fracture toughness reduces due to thermal embrittlement, some aged CASS materials have the potential to have exceedingly low toughness. CASS can also show high toughness variability due to the variability of its microstructure. Recently an ASME Section XI Code Case N-838 has been proposed to evaluate the flaw tolerance based on probabilistic fracture mechanics (PFM). An assessment of mechanical-property degradation is an input to perform the flaw evaluation procedure in CASS components. There are at least four different models for predicting the change in J-R curves in CASS due to thermal aging. One model is proprietary and the other three are the Argonne/NUREG-CR/4513R1, the French/EDF and a Japanese model. In this work, two of the thermal aging models were reviewed, reproduced and validated against their example cases for each individual model. Both models were then utilized to assess the fully aged conditions for cases that covers a large spectrum of CASS J R curves with high COV (coefficient of variance). Finally, J-R curves distributions using both Argonne and French models were established by examining the actual chemical compositions of CASS materials found in some US PWR plants. The J-R curves distributions include 21 pipes/fittings in primary pipe loop as well as data from an EPRI report. The calculated toughness variability in a single LBB plant is compared using the Argonne and French models. Additionally the relationship of the “C” and “m” parameters used in the power-law J-R curve equations (J = C×Δam) was explored to determine the proper way to statistically vary the J-R curve in probabilistic analyses.

The Analytical and Experimental Research of In-Vessel Corium Retention in CPR1000

2010 ◽  
Author(s):  
Xing Chen ◽  
Shishun Zhang ◽  
Jiming Lin ◽  
Huiyong Zhang
Keyword(s):  
Experimental Research ◽  
Success Probability ◽  
Outer Wall ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Structural Failure ◽  
Pressurized Water ◽  
Thermal Failure ◽  
Near Future ◽  
Preliminary Phase

The analytical and experiment research of In-Vessel Corium Retention (IVR) in the Chinese Pressurized-water Reactor 1000 MWe (CPR1000) are introduced. The IVR research consists of preliminary phase and detailed phase. The analysis of thermal failure, structural failure and penetration failure of Reactor Pressure Vessel (RPV) and the experimental research of External Reactor Vessel Cooling (ERVC) are performed at preliminary phase. Analysis results show that the RPV failure is the dominated by thermal failure mode and the probability of the thermal failure is very low. Test results show that the IVR success probability for CPR1000 is about 99% if the Critical Heat Flux (CHF) of CPR1000 is the same as that of AP600. Further works, including the ERVC enhancement design, the CHF test of the RPV outer wall and the recalculation of the IVR success probability for CPR1000, will be performed at detailed phase in the near future.

Piping Stress-Strain Correlation for Seismic Loading

1988 ◽  
Vol 110 (4) ◽  
pp. 444-450
Author(s):  
G. Stawniczy ◽  
W. R. Bak ◽  
G. Hau
Keyword(s):  
Pressure Vessel ◽  
Failure Modes ◽  
Low Cycle Fatigue ◽  
Seismic Loading ◽  
Evaluation Procedure ◽  
Test Results ◽  
Cycle Fatigue ◽  
Stress Strain ◽  
Safety Margins ◽  
Fatigue Evaluation

This paper establishes limits on piping material strains for ASME Boiler and Pressure Vessel Code Level D loadings that ensure a limitation of deformation and provide suitable safety margins. In establishing the strain limits, potential piping failure modes due to compressive wrinkling and low-cycle fatigue are considered. A stress-strain correlation methodology to convert linear, elastically calculated Code Class 2 and 3 equation (9)-Level D stresses to strains is established. This correlation is based on the fatigue evaluation procedure of the Code and is verified by comparison with test results. A detailed discussion of test results compared with the stress-strain correlation methodology is also presented.

PKL Tests on Heterogeneous Inherent Boron Dilution Following SB-LOCA: Applicability to Reactor Scale

2010 ◽  
Author(s):  
Klaus Umminger ◽  
Simon Philipp Schollenberger ◽  
Se´bastien Cornille ◽  
Claire Agnoux ◽  
Delphine Quintin ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Limited Range ◽  
Test Facility ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Physical Phenomena ◽  
Boron Dilution

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

Flaw Evaluation Procedure for Cast Austenitic Stainless Steel Materials Using Thermal Aging Models

2017 ◽  
Author(s):  
M. F. Uddin ◽  
G. M. Wilkowski ◽  
S. Pothana ◽  
F. W. Brust
Keyword(s):  
Fracture Toughness ◽  
Thermal Aging ◽  
Austenitic Stainless Steels ◽  
Full Scale ◽  
Evaluation Procedure ◽  
Chemical Compositions ◽  
Large Variability ◽  
Acceptance Criterion ◽  
Operating Temperatures ◽  
Aging Models

Thermal embrittlement of cast austenitic stainless steels (CASS) can occur at reactor operating temperatures potentially leading to a reduction in their fracture toughness. Some aged CASS materials have the potential to have exceedingly low toughness and also show high toughness variability due to the nature of their microstructure. The experimentally measured JIc values for CASS materials showed a large scatter when plotted against ferrite number (FN) or chrome equivalent number (Creq). Because of their low aged toughness with such a large variability, flaw evaluations of CASS material needs to be done carefully, especially since most US PWR nuclear plants have been given plant-life extensions for 60-year operation, and consideration of further extension to 80 years is underway. However, the ASME Section XI Appendix C flaw acceptance criterion currently does not have a recommended procedure for flaw evaluation for CASS materials with FN ≥ 20, and the Working Group recognizes that the changes might also be needed for CASS with FN less than 20. In this paper, a flaw evaluation procedure for fully aged CASS materials is presented using JIc values at LWR operating temperatures predicted from several existing thermal-aging toughness degradation models. All available thermal aging models for CASS materials were evaluated which predict fully aged (lower saturated toughness condition) fracture toughness of CASS based on their chemical compositions. A set of 20 experimental test data was analyzed by using all models to find the most accurate thermal aging models. Using the most accurate models, correlations between predicted JIc values and French Creq-Fr and ASTM A800 FN were developed from a database of 274 pipe/elbows in US PWR plants whose chemical compositions were known. Finally, the correlation was used to determine the elastic-plastic fracture correction factor (Z factor) for CASS pipe and fittings as a function of pipe diameter and their chemical compositions from material certification sheet using the Dimensionless-Plastic-Zone-Parameter (DPZP) analysis. The DPZP analysis is a relatively simple curve-fitting procedure through full-scale circumferential surface-cracked pipe tests developed in pipe fracture projects funded by the USNRC, and was checked against a full-scale aged CF8m pipe fracture test. After determining the chemical composition specific Z factor for CASS materials, the flaw evaluation can be performed according to the ASME Section XI Appendix C procedures.

scholarly journals Installation of Valvtechnologies Ball Valves in Severe Service Applications Within the Advanced Test Reactor

2006 ◽  
Author(s):  
Robert C. Howard
Keyword(s):  
National Laboratory ◽  
The United States ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Maintenance ◽  
Design Requirements ◽  
Test Reactor ◽  
Reactor Operating ◽  
Design Power ◽  
Operating Cycles

The Advanced Test Reactor (ATR) located at the Department of Energy’s Idaho National Laboratory, is the most powerful test reactor operating in the United States rated at a design power of 250 MW(t). Operating cycles are nominally seven per year with outages that last 7 to 14 days, allowing time for routine plant maintenance and experiment insertions and manipulations. While the ATR pressurized water loops can operate at the same temperature and pressure requirements of a pressurized water reactor, the loops also have the ability to operate at higher conditions. Hence, it is critical to ensure that when component replacements are called for, they can meet or exceed design requirements of a typical power reactor, while continuing to satisfy the design requirements of the ATR experiment loops.

A Study of Cavitation Erosion

2002 ◽  
Author(s):  
Hiromu Isaka ◽  
Masatsugu Tsutsumi ◽  
Hiroyuki Kobayashi ◽  
Tadashi Shiraishi
Keyword(s):  
Stainless Steel ◽  
Erosion Rate ◽  
Cavitation Erosion ◽  
Design Criteria ◽  
Hard Material ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Type Stainless Steel ◽  
Critical Cavitation

The authors performed experimental study for the purpose of the following two items from a viewpoint of cavitation erosion of a cylindrical orifice in view of a problem at the letdown orifice in PWR (Pressurized Water Reactor). 1. To get the critical cavitation parameter of the cylindrical orifice to establish the design criteria for prevention of cavitation erosion, and 2. to ascertain the erosion rate in such an eventuality that the cavitation erosion occurs with the orifice made of stainless steel with precipitation hardening (17-4-Cu hardening type stainless steel), so that we confirm the appropriateness of the design criteria. Regarding the 1st item, we carried out the cavitation tests to get the critical cavitation parameters inside and downstream of the orifice. The test results showed that the cavitation parameter at inception is independent of the length or the diameter of the orifice. Moreover, the design criteria of cavitation erosion of cylindrical orifices have been established. Regarding the 2nd item, we tested the erosion rate under high-pressure conditions. The cavitation erosion actually occurred in the cylindrical orifice at the tests that was strongly resemble to the erosion occurred at the plant. It will be seldom to reproduce resemble cavitation erosion in a cylindrical orifice with the hard material used at plants. We could establish the criteria for preventing the cavitation erosion from the test results.

Technical Basis for Revision of Code Case N-809 on Reference Fatigue Crack Growth Curves for Austenitic Stainless Steels in Pressurized Water Reactor Environments

2021 ◽  
Author(s):  
Russell C. Cipolla ◽  
Warren H. Bamford ◽  
Kiminobu Hojo ◽  
Yuichiro Nomura
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Stainless Steels ◽  
Growth Curves ◽  
Austenitic Stainless Steels ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
R Ratio ◽  
Technical Basis

Abstract Reference fatigue crack growth curves for austenitic stainless steels exposed to pressurized water reactor environments have been available in the ASME Code, Section XI in their present form with the publication of Code Case N-809 in Supplement 2 to the 2015 Code Edition. The reference curves are dependent on temperature, loading rate (loading rise time), mean stress (R-ratio), and cyclic stress intensity factor range (ΔK), which are all contained in the model. Since the first implementation of this Code Case, additional data have become available, and the purpose of this paper is to provide the technical basis for revision of the Code Case. Changes have been made in three areas: R-ratio behavior, threshold for crack growth (ΔKth), and crack growth rate dependence on ΔK. In addition, the temperature model was revisited to study the temperature effects for T < 150°C, where the current model predicts an increase in da/dN based on limited test data at about 100°C (200°F). At this point, the current temperature model is considered conservative and no change is proposed in this revision to N-809. The R-ratio model has been revised for both high and low carbon stainless steels, a significant improvement over the original procedures. Perhaps the most important revision is in the area of the threshold for the initiation of fatigue crack growth; such data are difficult to obtain, and the previous model was very conservative. Finally, the crack growth exponent was revised slightly to make it consistent with the regression analysis of the original data.

Sign in / Sign up

Export Citation Format

Share Document