13 Matching Results

Search Results

Advanced search parameters have been applied.

Results of irradiated cladding tests and clad plate experiments

Description: Two aspects critical to the fracture behavior of three-wire stainless steel cladding were investigated by the Heavy-Section Steel Technology (HSST) Program: (1) radiation effects on cladding strength and toughness, and (2) the response of mechanically loaded, flawed structures in the presence of cladding (clad plate experiments). Postirradiation testing results show that, in the test temperature range from /minus/125 to 288/degree/C, the yield strength increased, and ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing. Radiation damage decreased the Charpy upper-shelf energy by 15 to 20% and resulted in up to 28/degree/C shifts of the Charpy impact transition temperature. Results of irradiated 12.5-mm-thick compact specimens (0.5TCS) show consistent decreases in the ductile fracture toughness, J/sub Ic/, and the tearing modulus. Results from clad plate tests have shown that (1) a tough surface layer composed of cladding and/or heat-affected zone has arrested running flaws under conditions where unclad plates have ruptured, and (2) the residual load-bearing capacity of clad plates with large subclad flaws significantly exceeded that of an unclad plate. 13 figs., 1 tab.
Date: January 1, 1988
Creator: Haggag, F.M. & Iskander, S.K.
Partner: UNT Libraries Government Documents Department

Effects of irradiation on initiation and crack-arrest toughness of two high-copper welds and on stainless steel cladding

Description: The objective of the study on the high-copper welds is to determine the effect of neutron irradiation on the shift and shape of the ASME K{sub Ic} and K{sub Ia} toughness curves. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Compact specimens fabricated from these welds were irradiated at a nominal temperature of 288{degree}C to fluences from 1.5 to 1.9 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV). The fracture toughness test results show that the irradiation-induced shifts at 100 MPa/m were greater than the Charpy 41-J shifts by about 11 and 18{degree}C. Mean curve fits indicate mixed results regarding curve shape changes, but curves constructed as lower boundaries to the data do indicate curves of lower slopes. A preliminary evaluation of the crack-arrest results shows that the neutron-irradiation induced crack-arrest toughness temperature shift is about the same as the Charpy V-notch impact temperature shift at the 41-J energy level. The shape of the lower bound curves (for the range of test temperatures covered), compared to those of the ASME K{sub Ia} curve did not appear to have been altered by the irradiation. Three-wire stainless steel weld overlay cladding was irradiated at 288{degree}C to fluences of 2 and 5 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV). Charpy 41-J temperature shifts of 13 and 28{degree}C were observed, respectively. For the lower fluence only, 12.7-mm thick compact specimens showed decreases in both J{sub Ic} and the tearing modulus. Comparison of the fracture toughness results with typical plate and a low upper-shelf weld reveals that the irradiated stainless steel cladding possesses low ductile initiation fracture toughness comparable to the low upper-shelf weld. 8 refs., 12 figs., 2 tabs.
Date: January 1, 1990
Creator: Nanstad, R.K.; Iskander, S.K. & Haggag, F.M.
Partner: UNT Libraries Government Documents Department

Radiation-induced temperature shift of thhe ASME K/sub Ic/ curve

Description: The objective of this study was to determine the effects of neutron irradiation on the temperature shift and shape of the K/sub Ic/ curve described in Sect. XI of the ASME Boiler and pressure Vessel Code. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 215-mm-thick plate. Charpy impact, tensile, dropweight, and compact specimens up to 203.2 mm thick were fabricated and tested to provide a large data for unirradiated material. Similar specimens with compacts up to 101.6 mm thick, irradiated at about 288/degree/C to a mean fluence of about 1.6 /times/ 10/sup 19/ neutrons/cm/sup 2/ in the Oak Ridge Research Reactor, were tested to provide a similarly large data base with which to evaluate the temperature shift and shape of the ASME K/sub Ic/ curves. Testing was performed by both Oak Ridge National Laboratory and Materials Engineering Associates. Both linear-elastic and elastic-plastic fracture mechanics techniques were used to analyze test results. 3 refs., 4 figs., 1 tab.
Date: January 1, 1989
Creator: Nanstad, R.K.; Haggag, F.M. & Iskander, S.K.
Partner: UNT Libraries Government Documents Department

Effects of irradiation on strength and toughness of commercial LWR vessel cladding

Description: The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the three-wire series-arc commercial method. Cladding was applied in three layers to provide adequate thickness for the fabrication of test specimens. The three-wire series-arc procedure, developed by Combustion Engineering, Inc., Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. Charpy V-notch and tensile specimens were irradiated at 288/sup 0/C to fluence levels of 2 and 5 x 10/sup 19/ neutrons/cm/sup 2/ (>1 MeV). Postirradiation testing results show that, in the test temperature range from -125 to 288/sup 0/C, the yield strength increased by 8 to 30%, ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced, due to irradiation exposure, 15 and 20%, while the lateral expansion was reduced 43 and 41%, at 2 and 5 x 10/sup 19/ neutrons/cm/sup 2/ (>1 MeV), respectively. In addition, radiation damage resulted in 13 and 28/sup 0/C shifts of the Charpy impact transition temperature at the 41-J level for the low and high fluences, respectively.
Date: January 1, 1987
Creator: Haggag, F.M.; Corwin, W.R.; Alexander, D.J. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Materials property testing using a stress-strain microprobe

Description: The Stress-Strain Microprobe (SSM) uses an automated ball indentation technique to obtain flow data from a localized region of a test specimen or component. This technique is used to rapidly determine the yield strength and microstructural condition of a variety of materials including pressure vessel steels, stainless steels, and nickel-base alloys. The SSM provides an essentially non-destructive technique for the measurement of yield strength data. This technique is especially suitable for the study of complex or highly variable microstructures such as weldments and weld heat affected zones. In this study 119 distinct SSM determinations of the yield strength of eight engineering alloys are discussed and compared to data obtained by conventional tensile tests. The sensitivity of the SSM to the presence of residual stresses is also discussed.
Date: September 1, 1998
Creator: Panayotou, N.F.; Baldrey, D.G. & Haggag, F.M.
Partner: UNT Libraries Government Documents Department

Effects of thermal aging and neutron irradiation on the mechanical properties of three-wire stainless steel weld overlay cladding

Description: Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288{degrees}C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect of aging and neutron irradiation at 288{degrees}C to a fluence of 5 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) was a 22% reduction in the USE and a 29{degrees}C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to {minus}125{degrees}C) but no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging alone. Irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. Other results from tensile, CVN, and fracture toughness specimens showed that the effects of thermal aging at 288 or 343{degrees}C for 20,000 h each were very small and similar to those at 288{degrees}C for 1605 h. The effects of long-term thermal exposure time (50,000 h and greater) at 288{degrees}C will be investigated as the specimens become available in 1996 and beyond.
Date: May 1, 1997
Creator: Haggag, F.M. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

The use of field indentation microprobe in measuring mechanical properties of welds

Description: A field indentation microprobe (FIM) was conceived for evaluating the structural integrity of metallic components (including base metal, welds, and heat-affected zones) in situ in a nondestructive manner. The FIM consists of an automated ball indentation (ABI) unit for determining the mechanical properties (yield strength, flow properties, estimates of fracture toughness, etc.) and a nondestructive evaluation (NDE) unit (consisting of ultrasonic transducers and a video camera) for determining the physical properties such as crack size, material pileup around indentation, and residual stress presence and orientation. The laboratory version used in this work performs only ABI testing. ABI tests were performed on stainless steel base metal (type 316L), heat-affected zone, and welds (type 308). Excellent agreement was obtained between yield strength and flow properties (true-stress/true-plastic-strain curve) measured by the ABI tests and those from uniaxial tensile tests conducted on 308 stainless steel welds, thermally aged at 343/degree/C for different times, and on the base material. 4 refs., 17 figs.
Date: January 1, 1989
Creator: Haggag, F.M.; Wong, H.; Alexander, D.J. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Effects of irradiation on K/sub Ic/ curves for high-copper welds

Description: The Fifth Irradiation Series in the Heavy-Section Steel Technology (HSST) Program is aimed at obtaining a statistically significant fracture toughness data base on two weldments with high-copper contents to determine the shift and shape of the K/sub Ic/ curve as a consequence of irradiation. The program includes irradiated Charpy V-notch impact, tensile, and drop-weight specimens in addition to compact fracture toughness specimens. Compact specimens (CS) with thicknesses of 25.4, 50.8, and 101.6 mm (1TCS, 2TCS, and 4TCS, respectively) have been irradiated. Additionally, unirradiated 6TCS and 8TCS have been tested to attain the same K/sub Ic/ measuring capacity as the irradiated specimens. The materials for this irradiation series are two weldments fabricated from special heats of weld wire with copper added to the melt. One lot of Linde 0124 flux was used for all the welds. Copper levels for the two welds are 0.23 and 0.31 wt %, while the nickel contents are 0.60 wt %. 17 refs., 16 figs., 9 tabs.
Date: January 1, 1988
Creator: Nanstad, R.K.; McCabe, D.E.; Menke, B.H.; Iskander, S.K. & Haggag, F.M.
Partner: UNT Libraries Government Documents Department

Statistical analyses of fracture toughness results for two irradiated high-copper welds

Description: The objectives of the Heavy-Section Steel Irradiation Program Fifth Irradiation Series were to determine the effects of neutron irradiation on the transition temperature shift and the shape of the K{sub Ic} curve described in Sect. 6 of the ASME Boiler and Pressure Vessel Code. Two submerged-arc welds with copper contents of 0.23 and 0.31% were commercially fabricated in 215-mm-thick plates. Charpy V-notch (CVN) impact, tensile, drop-weight, and compact specimens up to 203.2 mm thick (1T, 2T, 4T, 6T, and 8T C(T)) were tested to provide a large data base for unirradiated material. Similar specimens with compacts up to 4T were irradiated at about 288{degrees}C to a mean fluence of about 1.5 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV) in the Oak Ridge Research Reactor. Both linear-elastic and elastic-plastic fracture mechanics methods were used to analyze all cleavage fracture results and local cleavage instabilities (pop-ins). Evaluation of the results showed that the cleavage fracture toughness values determined at initial pop-ins fall within the same scatter band as the values from failed specimens; thus, they were included in the data base for analysis (all data are designated K{sub Jc}).
Date: January 1, 1990
Creator: Nanstad, R.K.; McCabe, D.E.; Haggag, F.M.; Bowman, K.O. & Downing, D.J.
Partner: UNT Libraries Government Documents Department

Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

Description: The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low (121 to 210{degrees}C (250--410{degrees}F)) compared to those for commercial light-water reactors (LWRs) ({approximately}288{degrees}C (550{degrees}F)). The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 {times}10{sup 18} neutron/cm{sup 2} (0.68 {times} 10{sup 18} neutron/cm{sup 2} (>1 MeV)) at temperatures of 288, 204, 163, and 121{degrees}C (550, 400, 325, and 250{degrees}F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs.
Date: January 1, 1991
Creator: Haggag, F.M.; Nanstad, R.K. (Oak Ridge National Lab., TN (United States)) & Byrne, S.T. (ABB/Combustion Engineering, Inc., Windsor, CT (United States))
Partner: UNT Libraries Government Documents Department

Irradiation effects on fracture toughness of two high-copper submerged-arc welds, HSSI series 5. Volume 2, Appendices E and F

Description: The Fifth Irradiation Series in the Heavy-Section Steel irradiation (HSSI) Program was aimed at obtaining a statistically significant fracture toughness data base on two weldments with high-copper contents to determine the shift and shape of the K{sub lc} curve as a consequence of irradiation. The program included irradiated Charpy V-notch impact, tensile, and drop-weight specimens in addition to compact fracture toughness specimens. Compact specimens with thicknesses of 25.4, 50.8, and 101.6 mm [1T C(T), 2T C(T), and 4T C(T), respectively] were irradiated. Additionally, unirradiated 6T C(T) and 8T C(T) specimens with the same K{sub lc} measuring capacity as the irradiated specimens were tested. The materials for this irradiation series were two weldments fabricated from special heats of weld wire with copper added to the melt. One lot of Linde 0124 flux was used for all the welds. Copper levels for the two welds are 0.23 and 0.31 wt %, while the nickel contents for both welds are 0.60 wt %. Twelve capsules of specimens were irradiated in the pool-side facility of the Oak Ridge Research Reactor at a nominal temperature of 288{degree}C and an average fluence of about 1.5 {times} 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). This volume, Appendices E and F, contains the load-displacement curves and photographs of the fracture toughness specimens from the 72W weld (0.23 wt % Cu) and the 73 W weld (0.31 wt % Cu), respectively.
Date: October 1, 1992
Creator: Nanstad, R. K.; Haggag, F. M.; McCabe, D. E.; Iskander, S. K.; Bowman, K. O. & Menke, B. H.
Partner: UNT Libraries Government Documents Department

Aging management of major LWR components with nondestructive evaluation

Description: Nondestructive evaluation of material damage can contribute to continued safe, reliable, and economical operation of nuclear power plants through their current and renewed license period. The aging mechanisms active in the major light water reactor components are radiation embrittlement, thermal aging, stress corrosion cracking, flow-accelerated corrosion, and fatigue, which reduce fracture toughness, structural strength, or fatigue resistance of the components and challenge structural integrity of the pressure boundary. This paper reviews four nondestructive evaluation methods with the potential for in situ assessment of damage caused by these mechanisms: stress-strain microprobe for determining mechanical properties of reactor pressure vessel and cast stainless materials, magnetic methods for estimating thermal aging damage in cast stainless steel, positron annihilation measurements for estimating early fatigue damage in reactor coolant system piping, and ultrasonic guided wave technique for detecting cracks and wall thinning in tubes and pipes and corrosion damage to embedded portion of metal containments.
Date: December 31, 1997
Creator: Shah, V.N.; MacDonald, P.E.; Akers, D.W.; Sellers, C.; Murty, K.L.; Miraglia, P.Q. et al.
Partner: UNT Libraries Government Documents Department

Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

Description: The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab.
Date: July 1, 1995
Creator: Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E. et al.
Partner: UNT Libraries Government Documents Department