In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320°C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of ≈5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking. iv Intentionally Left BlankForeword v This report presents crack growth rate data and the results of the corresponding fracture surface and metallographic examinations from cyclic loading and primary water stress-corrosion cracking (PWSCC) tests of two nickel-base Alloy 182 (A182) weldments, which are typical of those used in vessel penetrations and piping butt welds in nuclear power plants. The effect of crack orientation with respect to dendrite orientation is the most significant variable investigated in this study. However, this report also includes a review of data from several laboratories, which describes the effects of material composition, loading characteristics, and chemistry of the aqueous environment. The main conclusion is that the PWSCC growth rates described for A182 specimens in this report are comparable to the crack growth rates that characterize the performance of Alloy 600 (A600).This report is the first in a series documenting the results of crack growth rate testing in vessel head penetration materials, focusing on the weld metals, A182 and A152, and including results of some tests of the base metals, A600 and (eventually) A690. The results presented in this report were obtained in tests of a laboratory-fabricated, shielded metal arc welding deposit of A182. ...

A procedure and correlations are presented for assessing thermal embrittlement and predicting Charpy-impact energy and fracture toughness J-R curve of cast stainless steel components under light water reactor operating conditions from known material information. The "saturation" impact strength and fracture toughness of a specific cast stainless steel, i.e., the minimum value that would be achieved for the material after long-term service, is estimated from the chemical composition of the steel. Fracture properties as a function of time and temperature of reactor service are estimated from the kinetics of embrittlement, which are also determined from chemical composition. A common "predicted lower-bound" J-R curve for cast stainless steels of unknown chemical composition is also defined for a given grade of steel, ferrite content, and temperature. Examples of estimating fracture toughness of cast stainless steel components during reactor service are presented.

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 ≈13 y at ≈281°C (538°F) for the hot-leg components and ≈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 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 ≈15 y and the KRB reactor pump cover plate (CF-8) after ≈8 y of service.

The following documents In the NUREG series are available for purchase from the Government Printing Office:formal NRC staff and contractor reports, NRC-sponsored conference proceedings, international agreement reports, grantee reports, and NRC booklets and brochures. Also available a r e regulatory guides, NRC regula- Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensileloading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high-and commercial-purity Type 304 SS specimens from controlblade absorber tubes irradiated in boiling water reactors were determined in slow-strain-ratetensile tests at 288OC. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.