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Measured residual stresses in overlay pipe weldments removed from service

Description: Surface and throughwall residual stresses were measured on an elbow-to-pipe weldment that had been removed from the Hatch-2 reactor about a year after the application of a weld overlay. The results were compared with experimental measurements on three mock-up weldments and with finite-element calculations. The comparison shows that there are significant differences in the form and magnitude of the residual stress distributions. However, even after more than a year of service, the residual stresses over most of the inner surface of the actual plant weldment with an overlay were strongly compressive. 3 refs., 7 figs.
Date: February 1, 1985
Creator: Shack, W.J.
Partner: UNT Libraries Government Documents Department

An overview of environmental degradation of materials in nuclear power plant piping systems

Description: Piping in light water reactor (LWR) power systems is affected by several types of environmental degradation: intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping in boiling water reactors (BWRs) has required research, inspection, and mitigation programs that will ultimately cost several billion dollars; erosion-corrosion of carbon steel piping has been observed frequently in the secondary systems of both BWRs and pressurized water reactors (PWRs); the effect of the BWR environment can greatly diminish the design margin inherent in the ASME Section III fatigue design curves for carbon steel piping; and cast stainless steels are subject to embrittlement after extended thermal aging at reactor operating temperatures. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions.
Date: August 1, 1987
Creator: Shack, W.J.
Partner: UNT Libraries Government Documents Department

Effects of nominal and crack-tip strain rate on IGSCC susceptibility in CERT tests

Description: Constant extension rate tests have been performed on sensitized Type 316 stainless steel in oxygenated water (8 ppM O/sub 2/) containing chloride ion impurities (0.5 ppM) over a range of strain rates from 10/sup -5/ to 2 x 10/sup -7/ s/sup -1/. The susceptibility to IGSCC (as quantified by parameters such as crack length at failure) increases with a decrease in strain rate. A model consistent with the observed and postulated crack growth behavior and with a fracture criterion is presented and used to derive power laws that relate the IGSCC susceptibility parameters and strain rate. The predicted strain rate exponents are in agreement with the experimental results of this and other studies. The correlations between IGSCC susceptibility and strain rate can be used to predict susceptibility to cracking outside the range of conditions used in the laboratory. In addition, it is shown that the average crack-tip strain rate in CERT experiments can be estimated by use of a J-integral approach. It is observed that the average crack growth rate is proportional to the square root of the estimated average crack-tip strain rate. The experimentally observed correlation is in good agreement with that deduced from a slip-dissolution model proposed by Ford.
Date: September 1, 1983
Creator: Maiya, P.S. & Shack, W.J.
Partner: UNT Libraries Government Documents Department

Leak rate measurements and detection systems

Description: A research program is under way to evaluate and develop improve leak detection systems. The primary focus of the work has been on acoustic emission detection of leaks. Leaks from artificial flaws, laboratory-generated IGSCCs and thermal fatigue cracks, and field-induced intergranular stress corrosion cracks (IGSCCs) from reactor piping have been examined. The effects of pressure, temperature, and leak rate and geometry on the acoustic signature are under study. The use of cross-correlation techniques for leak location and pattern recognition and autocorrelation for source discrimination is also being considered.
Date: October 1, 1983
Creator: Kupperman, D.; Shack, W.J. & Claytor, T.
Partner: UNT Libraries Government Documents Department

Intergranular crack propagation rates in sensitized Type 304 stainless steel in an oxygenated water environment

Description: Intergranular stress-corrosion crack (IGSCC) propagation rates were measured in three heats of sensitized Type 304 stainless steel (SS) as a function of applied load and sensitization in high-purity water with 8 ppM. Active-loading tests yielded IGSCC propagation rates ranging from approx. 2 x 10/sup -10/ to 1 x 10/sup -9/ m/s (approx. 2 x 10/sup -5/ to 2 x 10/sup -4/ in./h) over the range of stress intensities from 25 to 46 MPa..sqrt..m (22 to 41 ksi..sqrt..in.). If the dependence of propagation rate on stress intensity is assumed to follow a power law, a least-squares fit of data yields (da/dt) = 1.23 x 10/sup -8/ K/sup 2/ /sup 42/ (in./h) for K in ksi..sqrt..in. Deflection-controlled tests on standard 12.7-mm-thick compact tension specimens yielded IGSCC propagation rates from 7 x 10/sup -12/ to 2 x 10/sup -10/ m/s (10/sup -6/ to 2 x 10/sup -5/ in./h) at effective average stress intensities in the range 21 to 26 MPa..sqrt..m (19 to 24 ksi..sqrt..in.). Crack lengths were determined by compilance measurements using in-situ high-temperature clip gage or LVDT methods, optical metallography on the side faces of the specimen, and fractography of the cracked surface after completion of the tests. The optical metallography measurements did not provide useful estimates of crack lengths, because large variations in IGSCC propagation across the thickness of the specimens occurred. The effects of the degree of sensitization on the IGSCC propagation rate are obscured by the data scatter. However, it seems clear that these variables do not lead to order-of-magnitude changes in the crack propagation rate.
Date: December 1, 1983
Creator: Park, J.Y. & Shack, W.J.
Partner: UNT Libraries Government Documents Department

Mechanical properties of thermally aged cast stainless steels from shippingport reactor components.

Description: Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was {approx}13 y at {approx}281 C (538 F) for the hot-leg components and {approx}264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and JIC of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of {approx}15 y and the KRB reactor pump cover plate (CF-8) after {approx}8 y of service.
Date: June 7, 1995
Creator: Chopra, O. K.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Review of environmental effects on fatigue crack growth of austenitic stainless steels.

Description: Fatigue and environmentally assisted cracking of piping, pressure vessel cladding, and core components in light water reactors are potential concerns to the nuclear industry and regulatory agencies. The degradation processes include intergranular stress corrosion cracking of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or stress corrosion cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Crack growth data for wrought and cast austenitic SSs in simulated BWR water, developed at Argonne National Laboratory under US Nuclear Regulatory Commission sponsorship over the past 10 years, have been compiled into a data base along with similar data obtained from the open literature. The data were analyzed to develop corrosion-fatigue curves for austenitic SSs in aqueous environments corresponding to normal BWR water chemistries, for BWRs that add hydrogen to the feedwater, and for pressurized water reactor primary-system-coolant chemistry. The corrosion-fatigue data and curves in water were compared with the air line in Section XI of the ASME Code.
Date: July 11, 1994
Creator: Shack, W. J.; Kassner, T. F. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Intergranular Crack Propagation Rates in Sensitized Type 304 Stainless Steel in an Oxygenated Water Environment

Description: Intergranular stress-corrosion crack (IGSCC) propagation rates were measured in three heats of sensitized Type 304 stainless steel (SS) as a function of sensitization in an environment of high-purity water with 8 ppm oxygen, using a fracture mechanics approach. Specimens were sensitized using controlled furnace heat treatments and the degree of sensitization was measured by the electrochemical potentiokinetic reactivation (EPR) method. Active loading tests were performed on standard specimens over a range of intensities. Crack lengths were determined by compilance measurements using in-situ high-temperature clip gage or LVDT methods, optical metallography on the side faces of the specimen, and fractography of the cracked surface after completion of the tests. The optical metallography measurements did not provide useful estimates of crack lengths, because large variations in IGSCC propagation across the thickness of the specimens occurred. The effects of the degree of sensitization on the IGSCC propagation rate are obscured by the data scatter. However, it seems clear that these variables do not lead to order-of-magnitude changes in the crack propagation rate.
Date: December 1, 1983
Creator: Park, J. Y. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Stress corrosion cracking of candidate materials for nuclear waste containers

Description: Types 304L and 316L stainless steel (SS), Incoloy 825, Cu, Cu-30%Ni, and Cu-7%Al have been selected as candidate materials for the containment of high-level nuclear waste at the proposed Yucca Mountain Site in Nevada. The susceptibility of these materials to stress corrosion cracking has been investigated by slow-strain-rate tests (SSRTs) in water which simulates that from well J-13 (J-13 water) and is representative of the groundwater present at the Yucca Mountain site. The SSRTs were performed on specimens exposed to simulated J-13 water at 93{degree}C and at a strain rate 10{sup {minus}7} s{sup {minus}1} under crevice conditions and at a strain rate of 10{sup {minus}8} s{sup {minus}1} under both crevice and noncrevice conditions. All the tests were interrupted after nominal elongation strains of 1--4%. Examination by scanning electron microscopy showed some crack initiation in virtually all specimens. Optical microscopy of metallographically prepared transverse sections of Type 304L SS suggests that the crack depths are small (<10 {mu}m). Preliminary results suggest that a lower strain rate increases the severity of cracking of Types 304L and 316L SS, Incoloy 825, and Cu but has virtually no effect on Cu-30%Ni and Cu-7%Al. Differences in susceptibility to cracking were evaluated in terms of a stress ratio, which is defined as the ratio of the increase in stress after local yielding in the environment to the corresponding stress increase in an identical test in air, both computed at the same strain. On the basis of this stress ratio, the ranking of materials in order of increasing resistance to cracking is: Types 304L SS < 316L SS < Incoloy 825 {congruent} Cu-30%Ni < Cu {congruent} Cu-7%Al. 9 refs., 12 figs., 7 tabs.
Date: September 1, 1989
Creator: Maiya, P.S.; Shack, W.J. & Kassner, T.F.
Partner: UNT Libraries Government Documents Department

Energy Technology Division research summary 2004.

Description: The Energy Technology (ET) Division provides materials and engineering technology support to a wide range of programs important to the US Department of Energy (DOE). The Division's capabilities are generally applied to technical issues associated with energy systems, biomedical engineering, transportation, and homeland security. Research related to the operational safety of commercial light water nuclear reactors (LWRs) for the US Nuclear Regulatory Commission (NRC) remains another significant area of interest for the Division. The pie chart below summarizes the ET sources of funding for FY 2004.
Date: May 6, 2004
Creator: Poeppel, R. B. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Fatigue crack initiation in carbon and low-alloy steels in light water reactor environments : mechanism and prediction.

Description: Section 111 of the ASME Boiler and Pressure Vessel Code specifies fatigue design curves for structural materials. The effects of reactor coolant environments are not explicitly addressed by the Code design curves. Recent test data illustrate potentially significant effects of light water reactor (LWR) coolant environments on the fatigue resistance of carbon and low-alloy steels. Under certain loading and environmental conditions, fatigue lives of test specimens may be shorter than those in air by a factor of {approx}70. The crack initiation and crack growth characteristics of carbon and low-alloy steels in LWR environments are presented. Decreases in fatigue life of these steels in high-dissolved-oxygen water are caused primarily by the effect of environment on growth of short cracks &lt; 100 {micro}m in depth. The material and loading parameters that influence fatigue life in LWR environments are defined. Fatigue life is decreased significantly when five conditions are satisfied simultaneously, viz., applied strain range, service temperature, dissolved oxygen in water, and S content in steel are above a threshold level, and loading strain rate is below a threshold value. Statistical models have been developed for estimating the fatigue life of these steels in LWR environments. The significance of the effect of environment on the current Code design curve is evaluated.
Date: January 27, 1998
Creator: Chopra, O. K. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Methods for incorporating the effects of LWR coolant environments in pressure vessel and piping fatigue evaluations.

Description: This paper summarizes the work performed at Argonne National Laboratory on the fatigue of piping and pressure vessel steels in the coolant environments of light water reactors. The existing fatigue strain vs. life ({var_epsilon}-N) data were evaluated to establish the effects of various material and loading variables, such as steel type, strain range, strain rate, temperature, and dissolved-oxygen level in water, on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves for carbon and low-alloy steels and austenitic stainless steels as a function of material, loading, and environment variables. Case studies of fatigue failures in nuclear power plants are presented, and the contribution of environmental effects to crack initiation is discussed. Methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the possible conservatism in the existing fatigue design curves of the ASME Code.
Date: July 31, 2002
Creator: Chopra, O. K. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Examination of overlay pipe weldments removed from the Hatch-2 reactor

Description: Laboratory ultrasonic examination (UT), dye penetrant examination (PT), metallography, and sensitization measurements were performed on Type 304 stainless steel overlay pipe weldments from the Hatch-2 BWR to determine the effectiveness of UT through overlays and the effects of the overlays on crack propagation in the weldments. Little correlation was observed between the results of earlier in-service ultrasonic inspection and the results of PT and destructive examination. Considerable difficulty was encountered in correctly detecting the presence of cracks by UT in the laboratory. Blunting of the crack tip by the weld overlay was observed, but there was no evidence of tearing or throughwall extension of the crack beyond the blunted region.
Date: February 1, 1985
Creator: Park, J.Y.; Kupperman, D.S. & Shack, W.J.
Partner: UNT Libraries Government Documents Department

Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented.
Date: March 1998
Creator: Chopra, O. K. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Overview of steam generator tube degradation and integrity issues

Description: The degradation of steam generator tubes in pressurized water nuclear reactors continues to be a serious problem. Primary water stress corrosion cracking is commonly observed at the roll transition zone at U-bends, at tube denting locations, and occasionally in plugs and sleeves. Outer-diameter stress corrosion cracking and intergranular attack commonly occur near the tube support plate crevice, near the tube sheet in crevices or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of circumferential cracking at the RTZ on both the primary and secondary sides. Segmented axial cracking at the tubes support plate crevices is also becoming more common. Despite recent advances in in-service inspection technology, a clear need still exists for quantifying and improving the reliability of in- service inspection methods with respect to the probability of detection of the various types of flaws and their accurate sizing. Improved inspection technology and the increasing occurrence of such degradation modes as circumferential cracking, intergranular attack, and discontinuous axial cracking have led to the formulation of a new performance-based steam generator rule. This new rule would require the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes perform the required safety function over the next operating cycle. The new steam generator rule will also be applied to severe accident conditions to determine the continued serviceability of a steam generator with degraded tubes in the event of a severe accident. Preliminary analyses are being performed for a hypothetical severe accident scenario to determine whether failure will occur first in the steam generator tubes, which would lead to containment bypass, or instead in the hot leg nozzle or surge line, which would not.
Date: October 1996
Creator: Diercks, D. R.; Shack, W. J. & Muscara, J.
Partner: UNT Libraries Government Documents Department

Mechanical properties of thermally aged cast stainless steels from Shippingport reactor components

Description: Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was {approximately}13 y at {approximately}281 C (538 F) for the hot-leg components and {approximately}264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 reactor hot and crossover-leg elbows (CF-8M steel) after service of {approximately} 15 y and the KRB reactor pump cover plate (CF-8) after {approximately} 8 y of service.
Date: April 1995
Creator: Chopra, O. K. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Tensile stress corrosion cracking of type 304 stainless steel irradiated to very high dose

Description: Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20--100 displacement per atom or dpa) by the end of life. The data bases and mechanistic understanding of, the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high dose, i.e., is it purely mechanical failure or is it stress-commotion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-11 reactor after irradiation to {approximately}50 dpa at {approximately}370 C. Slow-strain-rate tensile tests were conducted at 289 C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microcopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at low ECP, and this susceptibility led to poor work-hardening capability and low ductility.
Date: September 1, 2001
Creator: Chung, H. M.; Ruther, W. E.; Strain, R. V. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Evaluation of effects of LWR coolant environments on fatigue life of carbon and low-alloy steels

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figure I-90 of Appendix I to Section III of the Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Recent test data indicate a significant decrease in fatigue life of carbon and low-alloy steels in LWR environments when five conditions are satisfied simultaneously, viz., applied strain range, temperature, dissolved oxygen in the water, and sulfur content of the steel are above a minimum threshold level, and the loading strain rate is below a threshold value. Only a moderate decrease in fatigue life is observed when any one of these conditions is not satisfied. This paper summarizes available data on the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, and sulfur content on the fatigue life of carbon and low-alloy steels. The data have been analyzed to define the threshold values of the five critical parameters. Methods for estimating fatigue lives under actual loading histories are discussed.
Date: February 1, 1996
Creator: Chopra, O.K. & Shack, W.J.
Partner: UNT Libraries Government Documents Department

Effects of material and loading variables on fatigue life of carbon and low-alloy steels in LWR environments

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Section III of the Code specifies fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data suggest that the Code fatigue curves may not always be adequate in coolant environments. This paper reports the results of recent fatigue tests that examine the effects of steel type, strain rate, dissolved oxygen level, strain range, loading waveform, and surface morphology on the fatigue life of A106-Gr B carbon steel and A533-Gr B low-alloy steel in water.
Date: March 1, 1995
Creator: Chopra, O.K. & Shack, W.J.
Partner: UNT Libraries Government Documents Department

Effects of LWR coolant environments on fatigue S-N curves for carbon and low-alloy steels

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figure I-90 of Appendix I to Section III of the Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Recent test data indicate significant decreases in fatigue lives of carbon and low-alloy steels in LWR environments when five conditions are satisfied simultaneously: applied strain range, temperature, dissolved oxygen in the water, and S content of the steel are above minimum threshold levels, and loading strain rate is below a threshold value. Only moderate decrease in fatigue life is observed when any one of these conditions is not satisfied. This paper presents several methods that have been proposed for evaluating the effects of LWR coolant environments on fatigue S-N curves for carbon and low-alloy steels. Estimations of fatigue lives under actual loading histories are discussed.
Date: June 1, 1996
Creator: Chopra, O.K. & Shack, W.J.
Partner: UNT Libraries Government Documents Department

Crack growth behavior of candidate waste container materials in simulated underground water

Description: Fracture-mechanics crack growth tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L Stainless steel and Incoloy 825 at 93{degrees}C and 1 atmosphere of pressure in simulated J-13 well water, which is representative of the groundwater at the Yucca Mountain site in Nevada that is proposed for a high-level nuclear waste repository. Crack growth rates were measured under various load conditions: load ratios of 0.2--1.0, frequencies of 2 {times} 10{sup {minus}4}{minus}1 Hz, rise times of 1--5000 s, and peak stress intensities of 25--40 MPa{center_dot}m{sup {1/2}}. The measured crack growthrates are bounded by the predicted rates from the current ASME Section 11 correlation for fatigue crack growth rates of austenitic stainless steel in air. Environmentally accelerated crack growth was not evident in any of the three materials under the test conditions investigated.
Date: December 31, 1992
Creator: Park, J.Y.; Shack, W.J. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Effects of LWR environments on fatigue life of carbon and low-alloy steels

Description: SME Boiler and Pressure Vessel Code provides construction of nuclear power plant components. Figure I-90 Appendix I to Section III of the Code specifies fatigue design curves for structural materials. While effects of environments are not explicitly addressed by the design curves, test data suggest that the Code fatigue curves may not always be adequate in coolant environments. This paper reports the results of recent fatigue tests that examine the effects of steel type, strain rate, dissolved oxygen level, strain range, loading waveform, and surface morphology on the fatigue life of A 106-Gr B carbon steel and A533-Gr B low-alloy steel in water.
Date: March 1, 1995
Creator: Chopra, O.K. & Shack, W.J.
Partner: UNT Libraries Government Documents Department