20 research outputs found
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Stress corrosion cracking of candidate materials for nuclear waste containers
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
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Measured residual stresses in overlay pipe weldments removed from service
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
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An overview of environmental degradation of materials in nuclear power plant piping systems
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
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Evaluation of aging degradation of structural components
Irradiation embrittlement of the neutron shield tank (NST) A212 Grade B steel from the Shippingport reactor, as well as thermal embrittlement of CF-8 cast stainless steel components from the Shippingport and KRB reactors, has been characterized. Increases in Charpy transition temperature (CTT), yield stress, and hardness of the NST material in the low-temperature low-flux environment are consistent with the test reactor data for irradiations at < 232{degrees}C. The shift in CTT is not as severe as that observed in surveillance samples from the High Flux Isotope Reactor (HFIR): however, it shows very good agreement with the results for HFIR A212-B steel irradiated in the Oak Ridge Research Reactor. The results indicate that fluence rate has not effect on radiation embrittlement at rates as low as 2 {times} 10{sup 8} n/cm{sup 2}{center dot}s at the low operating temperature of the Shippingport NST, i.e., 55{degrees}C. This suggest that radiation damage in Shippingport NST and HFIR surveillance samples may be different because of the neutron spectra and/or Cu and Ni content of the two materials. Cast stainless steel components show relatively modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength. Correlations for estimating mechanical properties of cast stainless steels predict accurate or slightly conservative 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 predict the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of {approx}15 y
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Effects of nominal and crack-tip strain rate on IGSCC susceptibility in CERT tests
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
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Intergranular crack propagation rates in sensitized Type 304 stainless steel in an oxygenated water environment
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
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Embrittlement of the Shippingport reactor shield tank
The irradiation embrittlement of the Shippingport neutron tank material has been characterized. Irradiation increases the Charpy transition temperature (CTT) by {approximately}25{degrees}C (45{degrees}F) and decreases the upper shelf energy. The shift in CTT is not as severe as that observed in the HFIR surveillance specimens. However, the actual value of CTT is higher than that for the HFIR data and the toughness at service temperature is low. The increase in yield stress is 51 MPa (7.4 ksi), which is comparable to the HFIR data. The results also indicate a low impact strength and higher transition temperature for the TL orientation than that for the LT orientation. Some effects of the location across the thickness of the wall are also observed for the LT specimens; CTT is slightly greater for the specimens from the inner region of the wall. The data agree well with results from high-flux test reactors. Annealing studies indicate complete recovery of embrittlement after a 2-h anneal at 400{degrees}C. The transition curve for the annealed inner wall specimens is virtually identical to that for the as-received outer wall. The results for weld specimens from the inner and outer walls are also presented. 7 refs., 12 figs
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Leak rate measurements and detection systems
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
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Examination of overlay pipe weldments removed from the Hatch-2 reactor
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
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Stress corrosion crack growth rates in Type 304 stainless steel in simulated BWR environments
Stress corrosion cracking of Type 304 stainless steel has been studied with fracture-mechanics-type standard 25.4-mm-thick compact tension specimens in simulated boiling-water reactor environments at 289/sup 0/C and 8.3 MPa. Tests were performed with either constant or cyclic loading. The latter tests used a positive sawtooth waveform with an unloading time of 1 or 5 s, a load ratio R (minimum load to maximum load) of 0.2 to 0.95, and a frequency f of 8 x 10/sup -4/ to 1 x 10/sup -1/ Hz. Crack lengths and crack growth rates were determined by the compliance method; crack mouth opening displacement was measured with in-situ clip gauges. Fractography was used to examine the mode of cracking and to confirm the compliance method for crack length determination. The test environments were high-purity deionized water with 0.2- to 8-ppM dissolved oxygen, and water with 0.2-ppM dissolved oxygen and 0.1-ppM sulfate (as H/sub 2/SO/sub 4/). Two heats with a carbon content of 0.06 wt % were investigated in solution-heat-treated and furnace-sensitized conditions. Degree of sensitization varied from approx. 0 to 20 C/cm/sup 2/ as measured by the electrochemical potentiokinetic polarization method. 8 references, 7 figures, 5 tables