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Comparison between instrumented precracked Charpy and compact specimen tests of carbon steels

Description: The General Atomic Company High Temperature Gas-Cooled Reactor (HTGR) is housed within a prestressed concrete reactor vessel (PCRV). Various carbon steel structural members serve as closures at penetrations in the vessel. A program of testing and evaluation is underway to determine the need for reference fracture toughness (K/sub IR/) and indexing procedures for these materials as described in Appendix G to Section III, ASME Code for light water reactor steels. The materials of interest are carbon steel forgings (SA508, Class 1) and plates (SA537, Classes 1 and 2) as well as weldments of these steels. The fracture toughness behavior is characterized with instrumented precracked Charpy V-votch specimens (PCVN) - slow-bend and dynamic - and compact specimens (10-mm and 25-mm thicknesses) using both linear elastic (ASTM E399) and elastic-plastic (equivalent Energy and J-Integral) analytical procedures. For the dynamic PCVN tests, force-time traces are analyzed according to the procedures of the Pressure Vessel Research Council (PVRC)/Metal Properties Council (MPC). Testing and analytical procedures are discussed and PCVN results are compared to those obtained with compact specimens.
Date: January 1, 1980
Creator: Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Direct comparison of unloading compliance and potential drop techniques in J-integral testing

Description: Single-specimen J-integral testing is performed commonly with the unloading compliance technique. Use of modern instrumentation techniques and powerful desktop computers have made this technique a standard. However, this testing technique is slow and tedious, with the loading rate fixed at a slow quasi-static rate. For these reasons the dc potential drop technique was investigated for crack length measurement during a J-integral test. For direct comparison, both unloading compliance and potential drop were used simultaneously during a J-integral test. The results showed good agreement between the techniques. However, the potential drop technique showed an offset in crack length due to plastic blunting processes. Taking this offset into account, J/sub Ic/ values calculated by both techniques compared well.
Date: January 1, 1984
Creator: McGowan, J.J. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Observations on the behavior of surface flaws in the presence of cladding

Description: A small crack near the inner surface of clad nuclear reactor pressure vessels (RPV) is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, six clad and two unclad, were performed to determine the effect, if any, of stainless steel cladding upon the propagation of small surface cracks subjected to stress states similar to those produced by pressurized thermal shock conditions. Results of tests at temperature 10 and 60/degree/C below the nil-ductility-transition temperature (NDT) have shown that (1) a tough surface layer composed of cladding and/or heat-affected zone has arrested running flaws in clad plates 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. 6 refs., 7 figs., 2 tabs.
Date: January 1, 1988
Creator: Iskander, S.K. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Characterization of four prestressed concrete reactor vessel liner steels

Description: A program of fracture toughness testing and analysis is being performed with PCRV steels for HTGRs. This report focuses on background information for the base materials and results of characterization testing, such as tensile and impact properties, chemical composition, and microstructural examination. The steels tested were an SA-508 class 1 forging, two plates of SA-537 class 1, and one plate of SA-537 class 2. Tensile requirements in effect at the time of procurement are met by all four steels. However, the SA-537 class 2 plate would not meet the minimum requirement for yield strength. Drop-weight and Charpy impact tests verified that the RT/sub NDT/ is equal to the NDT for each steel. Charpy impact energies at the NDT range from 40 J (30 ft-lb) for one heat of SA-537 class 1 to 100 J (74 ft-lb) for the SA-537 class 2 plate; upper-shelf energies range from 170 to 310 J (125 to 228 ft-lb) for the same two steels, respectively. The onset of upper-shelf energy occurred at temperatures ranging from 0 to 50/sup 0/C.
Date: December 1, 1980
Creator: Nanstad, R. K.
Partner: UNT Libraries Government Documents Department

Fracture properties of a neutron-irradiated stainless steel submerged arc weld cladding overlay

Description: The ability of stainless steel cladding to increase the resistance of an operating nuclear reactor pressure vessel to extension of surface flaws depends greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the submerged arc, single-wire, oscillating-electrode method. Three layers of cladding provided a thickness adequate for fabrication of test specimens. The first layer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by excessive melting of the base plate. Specimens were taken from near the base plate-cladding interface and also from the upper layers. Charpy V-notch and tensile specimens were irradiated at 288/sup 0/C to a fluence of 2 x 10/sup 23/ neutrons/m/sup 2/ (>1 MeV). 10 refs., 16 figs., 4 tabs.
Date: January 1, 1984
Creator: Corwin, W.R.; Berggren, R.G. & Nanstad, R.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

The effect of aging at 343/degree/C on type 308 stainless steel welds

Description: Three nominally 25-mm (1-in) thick shielded metal-arc welds were prepared from 304L base plate with 308 filler material, to obtain three different ferrite levels (4, 8, and 12 %). Portions of these welds were then aged at 343/degree/C for 3000, 10000, and 20000 hours. Charpy V-notch and tensile specimens were taken from the welds. The tensile results were similar for all the specimens and showed little effect of aging on either the yield or ultimate tensile strengths. The Charpy impact properties of the higher ferrite content materials were significantly degraded by these agings, with larger decreases in the impact energy with increased aging time. The microstructures of the welds were examined by metallography and transmission electron microscopy, and the fracture surfaces were examined by scanning electron microscopy. The changes in the mechanical properties and the fractography are discussed in light of the observed changes in the microstructure. 3 refs., 3 figs., 2 tabs.
Date: January 1, 1989
Creator: Alexander, D.J.; Alexander, K.B. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Charpy toughness and tensile properties of a neutron irradiated stainless steel submerged-arc weld cladding overlay

Description: The possibility of stainless steel cladding increasing the resistance of an operating nuclear reactor pressure vessel to extension of surface flaws is highly dependent upon the irradiated properties of the cladding. Therefore, weld overlay cladding irradiated at temperatures and fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the submerged-arc, single-wire, oscillating electrode method. Three layers of cladding were applied to provide a cladding thickness adequate for fabrication of test specimens. The first layer was type 309, and the upper two layers were type 308 stainless steel. There was considerable dilution of the type 309 in the first layer of cladding as a result of excessive melting of the base plate. Specimens for the irradiation study were taken from near the base plate/cladding interface and also from the upper layers of cladding. Charpy V-notch and tensile specimens were irradiated at 288/sup 0/C to neutron fluences of 2 x 10/sup 23/ n/m/sup 2/ (E > 1 MeV). When irradiated, both types 308 and 309 cladding showed a 5 to 40% increase in yield strength accompanied by a slight increase in ductility in the temperature range from 25 to 288/sup 0/C. All cladding exhibited ductile-to-brittle transition behavior during impact testing.
Date: January 1, 1984
Creator: Corwin, W.R.; Berggren, R.G. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Post-irradiation Examination Plan for ORNL and University of California Santa Barbara Assessment of UCSB ATR-2 Irradiation Experiment

Description: New and existing databases will be combined to support development of physically based models of transition temperature shifts (TTS) for high fluence-low flux (φ < 10{sup 11}n/cm{sup 2}-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1×10{sup 20} n/cm{sup 2} (>1 MeV). All references to neutron flux and fluence in this report are for fast neutrons (>1 MeV). The reactor pressure vessel (RPV) task of the Light Water Reactor Sustainability (LWRS) Program is working with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural features from reactor surveillance materials with those from similar materials irradiated under high flux conditions in test reactors
Date: January 25, 2014
Creator: Nanstad, R. K.; Yamamoto, T. & Sokolov, M. A.
Partner: UNT Libraries Government Documents Department

Use of instrumented charpy tests to determine onset of upper-shelf energy

Description: The Charpy V-notch (C/sub v/) upper-shelf energy is usually defined as that temperature range in which the surface of the C/sub v/ specimen exhibits an appearance indicative of a 100 percent ductile fracture. In an attempt to avoid the need for interpretation, the selection of the C/sub v/ upper-shelf energy is based on the results from an instrumented impact test which provides a permanent record of the load-deflection history of a C/sub v/ specimen during the testing sequence. In the brittle-ductile transition temperature regime, a precipitous drop in the load trace occurs. The amount of the drop decreases at higher temperatures until it is zero, and the zero-drop-in-load temperature is identical to the onset of the C/sub v/ upper shelf. This relationship between the drop in load and energy in an instrumented impact test provides incontestable assurance that the C/sub v/ upper shelf has been obtained. This relationship between drop in load and temperature permits a prediction of the onset of the upper-shelf temperature with as few as two instrumented impact tests. It is also shown that nil-ductility temperature (NDT) (determined by the drop-weight test) is released to the C/sub v/ upper shelf. For the SA-508 Class 2 and SA-533 Grade B Class 1 steels employed in the fabrication of pressure vessels for light-water reactors, C/sub v/ testing at NDT + 180$sup 0$F (100$sup 0$C) will provide upper-shelf energy values. (DLC)
Date: November 1, 1975
Creator: Canonico, D.A.; Stelzman, W.J.; Berggren, R.G. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology program series 4 and 5)

Description: This report presents studies on the irradiation effects in low-alloy reactor pressure vessel steels. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (''current practice welds''). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds. 27 refs., 22 figs. (JDB)
Date: January 1, 1985
Creator: McGowan, J.J.; Nanstad, R.K.; Thoms, K.R. & Menke, B.H.
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

The effects of thermal annealing on fracture toughness of low upper-shelf welds

Description: Experimental results are presented from a study of the effects of thermal annealing on recovery of fracture toughness of low upper-shelf submerged-arc welds (weld designations 61W through 67W) from the Heavy-Section Steel Irradiation (HSSI) Program Second and Third Irradiation Series. Most of the study was conducted to evaluate the effects of annealing on the J-R curves of the submerged-arc welds. The recovery of fracture toughness in the transition range as the result of annealing was studied for welds 63W, 64, and 65W only. Compact specimens of 12.7- and 20.3-mm-thick (0.5T and 0.8T, respectively) were tested in this study. The specimens had been previously irradiated at the Oak Ridge National Laboratory (ORNL) Bulk Shielding Reactor. Each weld was irradiated to a certain value of neutron fluence in the range from 0.4 to 1.3 {times} 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the average temperature range of 275 to 300 C. Annealing of the irradiated specimens was done at 454 C for 168 h. Fracture toughness tests were performed at temperatures selected to match those of the previously conducted unirradiated and irradiated tests.
Date: December 31, 1994
Creator: Sokolov, M.A.; Nanstad, R.K. & Iskander, S.K.
Partner: UNT Libraries Government Documents Department

Effects of annealing time on the recovery of Charpy V-notch properties of irradiated high-copper weld metal

Description: One of the options to mitigate the effects of irradiation on reactor pressure vessels is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. An important issue to be resolved is the effect on the toughness properties of reirradiating a vessel that has been annealed. This paper describes the annealing response of irradiated high-copper submerged-arc weld HSSI 73W. For this study, the weld has been annealed at 454 C (850 F) for lengths of time varying between 1 and 14 days. The Charpy V-notch 41-J (30-ft-lb) transition temperature (TT{sub 41J}) almost fully recovered for the longest period studied, but recovered to a lesser degree for the shorter periods. No significant recovery of the TT{sub 41J} was observed for a 7-day anneal at 343 C (650 F). At 454 C for the durations studied, the values of the upper-shelf impact energy of irradiated and annealed weld metal exceeded the values in the unirradiated condition. Similar behavior was observed after aging the unirradiated weld metal at 460 and 490 C for 1 week.
Date: December 31, 1994
Creator: Iskander, S.K.; Sokolov, M.A. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

Description: One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.
Date: December 31, 1996
Creator: Nanstad, R.K.; Iskander, S.K. & Sokolov, M.A.
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

A perspective on thermal annealing of reactor pressure vessel materials from the viewpoint of experimental results

Description: It is believed that in the next decade or so, several nuclear reactor pressure vessels (RPVs) may exceed the reference temperature limits set by the pressurized thermal shock screening criteria. One of the options to mitigate the effects of irradiation on RPVs is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory to study the annealing response, or ``recovery`` of several irradiated RPV steels. The fracture toughness is one of the important properties used in the evaluation of the integrity of RPVs. Optimally, the fracture toughness is measured directly by fracture toughness specimens, such as compact tension or precracked Charpy specimens, but is often inferred from the results of Charpy V-notch impact specimens. The experimental results are compared to the predictions of models for embrittlement recovery which have been developed by Eason et al. Some of the issues in annealing that still need to be resolved are discussed.
Date: April 1, 1996
Creator: Iskander, S.K.; Sokolov, M.A. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Comparison of different experimental and analytical measures of the thermal annealing response of neutron-irradiated RPV steels

Description: The thermal annealing response of several materials as indicated by Charpy transition temperature (TT) and upper-shelf energy (USE), crack initiation toughness, K{sub Jc}, predictive models, and automated-ball indentation (ABI) testing are compared. The materials investigated are representative reactor pressure vessel (RPV) steels (several welds and a plate) that were irradiated for other tasks of the Heavy-Section Steel Irradiation (HSSI) Program and are relatively well characterized in the unirradiated and irradiated conditions. They have been annealed at two temperatures, 343 and 454 C (650 and 850 F) for varying lengths of time. The correlation of the Charpy response and the fracture toughness, ABI, and the response predicted by the annealing model of Eason et al. for these conditions and materials appears to be reasonable. The USE after annealing at the temperature of 454 C appears to recover at a faster rate than the TT, and even over-recovers (i.e., the recovered USE exceeds that of the unirradiated material).
Date: May 1, 1997
Creator: Iskander, S.K.; Sokolov, M.A. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Response of neutron-irradiated RPV steels to thermal annealing

Description: One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the fracture toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results of work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response of several irradiated RPV steels.
Date: March 1, 1997
Creator: Iskander, S.K.; Sokolov, M.A. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department

Charpy impact test results on five materials and NIST verification specimens using instrumented 2-mm and 8-mm strikers

Description: The Heavy-Section Steel Irradiation Program at Oak Ridge National Laboratory is involved in two cooperative projects, with international participants, both of which involve Charpy V-notch impact tests with instrumented strikers of 2mm and 8mm radii. Two heats of A 533 grade B class I pressure vessel steel and a low upper-shelf (LUS) submerged-arc (SA) weld were tested on the same Charpy machine, while one heat of a Russian Cr-Mo-V forging steel and a high upper-shelf (HUS) SA weld were tested on two different machines. The number of replicate tests at any one temperature ranged from 2 to 46 specimens. Prior to testing with each striker, verification specimens at the low, high, and super high energy levels from the National Institute of Standards and Technology (NIST) were tested. In the two series of verification tests, the tests with the 2mm striker met the requirements at the low and high energy levels but not at the super high energy. For one plate, the 2mm striker showed somewhat higher average absorbed energies than those for the 8-mm striker at all three test temperatures. For the second plate and the LUS weld, however, the 2mm striker showed somewhat lower energies at both test temperatures. For the Russian forging steel and the HUS weld, tests were conducted over a range of temperatures with tests at one laboratory using the 8mm striker and tests at a second laboratory using the 2mm striker. Lateral expansion was measured for all specimens and the results are compared with the absorbed energy results. The overall results showed generally good agreement (within one standard deviation) in energy measurements by the two strikers. Load-time traces from the instrumented strikers were also compared and used to estimate shear fracture percentage. Four different formulas from the European Structural Integrity Society draft standard for instrumented ...
Date: April 1, 1995
Creator: Nanstad, R.K. & Sokolov, M.A.
Partner: UNT Libraries Government Documents Department

Fracture toughness evaluation of a low upper-shelf weld metal from the Midland Reactor using the master curve

Description: The primary objective of the Heavy-Section Steel Irradiation (HSSI) Program Tenth Irradiation Series was to develop a fracture mechanics evaluation of weld metal WF-70, which was taken from the beltline and nozzle course girth weld joints of the Midland Reactor vessel. This material became available when Consumers Power Company of Midland, Michigan, decided to abort plans to operate their nuclear power plant. WF-70 is classified as a low upper-shelf steel primarily due to the Linde 80 flux that was used in the submerged-arc welding process. The master curve concept is introduced to model the transition range fracture toughness when the toughness is quantified in terms of K{sub Jc} values. K{sub Jc} is an elastic-plastic stress intensity factor calculated by conversion from J{sub c}; i.e., J-integral at onset of cleavage instability.
Date: March 1997
Creator: McCabe, D. E.; Sokolov, M. A. & Nanstad, R. K.
Partner: UNT Libraries Government Documents Department

On impact testing of subsize Charpy V-notch type specimens

Description: The potential for using subsize specimens to determine the actual properties of reactor pressure vessel steels is receiving increasing attention for improved vessel condition monitoring that could be beneficial for light-water reactor plant-life extension. This potential is made conditional upon, on the one hand, by the possibility of cutting samples of small volume from the internal surface of the pressure vessel for determination of actual properties of the operating pressure vessel. The plant-life extension will require supplemental surveillance data that cannot be provided by the existing surveillance programs. Testing of subsize specimens manufactured from broken halves of previously tested surveillance Charpy V-notch (CVN) specimens offers an attractive means of extending existing surveillance programs. Using subsize CVN type specimens requires the establishment of a specimen geometry that is adequate to obtain a ductile-to-brittle transition curve similar to that obtained from full-size specimens. This requires the development of a correlation of transition temperature and upper-shelf toughness between subsize and full-size specimens. The present study was conducted under the Heavy-Section Steel Irradiation Program. Different published approaches to the use of subsize specimens were analyzed and five different geometries of subsize specimens were selected for testing and evaluation. The specimens were made from several types of pressure vessel steels with a wide range of yield strengths, transition temperatures, and upper-shelf energies (USEs). Effects of specimen dimensions, including depth, angle, and radius of notch have been studied. The correlation of transition temperature determined from different types of subsize specimens and the full-size specimen is presented. A new procedure for transforming data from subsize specimens was developed and is presented.
Date: December 31, 1994
Creator: Mikhail, A.S. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department