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Two finite element techniques for computing mode I stress intensity factors in two- or three-dimensional problems

Description: Two finite element (FE) approaches were used to calculate opening mode I stress intensity factors (K/sub I/) in two- or three-dimensional (2-D and 3-D) problems for the Heavy-Section Steel Technology (HSST) program. For problems that can be modeled in two dimensions, two techniques were used. One of these may be termed an ''energy release rate'' technique, and the other is based on the classical near-tip displacement and stress field equations. For three-dimensional problems, only the latter technique was used. In the energy release technique, K/sub I/ is calculated as the change in potential energy of the structure due to a small change in crack length. The potential energy is calculated by the FE method but without completely solving the system of linear equations for the displacements. Furthermore, the system of linear equations is only slightly perturbed by the change in crack length and, therefore, many computations need not be repeated for the second structure with the slight change in crack length. Implementation of these last two items has resulted in considerable savings in the calculation of K/sub I/ as compared to two complete FE analyses. These ideas are incorporated in the FMECH code. The accuracy of the methods has been checked by comparing the results of the two approaches with each other and with closed form solutions. It is estimated that the accuracy of the results is about +-5%.
Date: February 1, 1981
Creator: Iskander, S.K.
Partner: UNT Libraries Government Documents Department

Method for LEFM analysis of RPV during SBLOCA

Description: A somewhat simplified two-dimensional linear elastic fracture mechanics model of the beltline of a RPV is presented. The fracture mechanics model used tends to be conservative in the sense that it ignores possible beneficial effects of warm prestressing and cladding. For LEFM studies that require a large number of analyses on the same geometry but with different loads and material toughnesses, the superposition principle is an accurate and simple method to determine K/sub 1/, provided that K/sub 1/ due to a unit load (called K* in this paper), acting on an arbitrary point on the crack surface is known. The details of the superposition principle and the procedure used for determining K* have been presented. It is believed that the error in the K/sub 1/ - values so determined is less than 3.5%. Once these K* - are determined for a specific geometry, then the determination of K/sub 1/ for the same geometry can be made accurately and in a manner that permits parametric studies involving thousands of individual analyses. An example of the use of the simplified model for a parametric analysis is also presented. 35 refs., 11 figs., 1 tab.
Date: January 1, 1985
Creator: Iskander, S.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

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

Inelastic analysis of two plates under deformation dependent loads

Description: Cover plates are used in current designs for high temperature gas-cooled reactors to compress the mineral fiber insulation against the inside of the liner of the prestressed concrete pressure vessel. In the upper plenum, these plates are hexagonal and specified as carbon steel; in the lower cross ducts, the plates are square and made of Hastelloy X. The General Atomic Company has specified both damage and safety limit criteria for these plates. These plates have been analyzed at these limits using the inelastic finite element computer program EPACA. The results indicate that the total strains for the square plate were within the specified values; however, the maximum deformations at the free corners indicate separation from the insulation and failure to achieve one of the design requirements. Since no material data were available for carbon steel at the limiting temperatures, it was assumed that the hexagonal plates were constructed of 2$sup 1$/$sub 4$ percent Cr--1 percent Mo material. Although this material was found to produce satisfactory performance, extrapolation of available information would lead to the conclusion that the performance of carbon steel plates would not be satisfactory at the specified conditions. (auth)
Date: February 1, 1976
Creator: Iskander, S.K.; Collins, C.W. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Review of PWR-related thermal-shock studies

Description: Flaw behavior trends associated with pressurized-thermal-shock (PTS) loading of PWR pressure vessels have been under investigation at ORNL for approx.12 years. During that time, eight thermal-shock experiments with thick-walled steel cylinders were conducted as a part of the investigations. These experiments demonstrated, in good agreement with linear elastic fracture mechanics (LEFM), crack initiation and arrest, a series of initiation-arrest events with deep penetration of the wall, long crack jumps without significant dynamic effects at arrest, arrest in a rising K/sub I/ field, extensive surface extension of an initially short and shallow flaw, and warm prestressing with K/sub I/ equal to or less than 0. This information was used in the development of a fracture-mechanics model that is being used extensively in the evaluation of the PTS issue.
Date: January 1, 1986
Creator: Cheverton, R.D.; Iskander, S.K. & Ball, D.G.
Partner: UNT Libraries Government Documents Department

Reirradiation Response Rate of a High-Copper Reactor Pressure Vessel Weld

Description: The Charpy impact response of reirradiated Heavy-Section Steel Irradiation (HSSI) Program Weld 73W has been determined at three fluence levels. The Charpy specimens had previously been irradiated at 288 C to 1.8 x 10{sup 19} cm{sup -2} (E > 1 MeV) and annealed at 454 C for 168 h. The results show that the change in the 41-J Charpy energy level transition temperature (OTT{sub 41-J}) of the reirradiated specimens is slightly higher than predicted by the vertical shift method, but significantly less than predicted by the lateral shift method. Previous results have also shown that the upper-shelf energy (USE) over-recovers as a consequence of annealing, which may explain why the USE value after a significant amount of reirradiation is approximately equal to the USE value in the unirradiated condition.
Date: October 22, 2001
Creator: Iskander, S. K.
Partner: UNT Libraries Government Documents Department

Integrity of PWR pressure vessels during overcooling accidents

Description: The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.
Date: January 1, 1982
Creator: Cheverton, R.D.; Iskander, S.K. & Whitman, G.D.
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

Integrity of PWR pressure vessels during overcooling accidents

Description: The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.
Date: January 1, 1982
Creator: Cheverton, R.D.; Iskander, S.K. & Whitman, G.D.
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

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

Results of charpy V-notch impact testing of structural steel specimens irradiated at {approximately}30{degrees}C to 1 x 10{sup 16} neutrons/cm{sup 2} in a commercial reactor cavity

Description: A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at {approximately} 30{degrees}C ({approximately} 85{degrees}F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 10{sup 16} neutrons/cm{sup 2} (> 1MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was {approximately} 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of {approximately} 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications.
Date: April 1997
Creator: Iskander, S. K. & Stoller, R. E.
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

Results of crack-arrest tests on irradiated a 508 class 3 steel

Description: Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K{sub la} of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10{degrees}C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTM A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280{degrees}C, and to a fluence varying from 1.7 to 2.7 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284{degrees}C to a fluence of 3.2 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K{sub la} curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements ...
Date: February 1, 1998
Creator: Iskander, S.K.; Milella, P.P. & Pini, M.A.
Partner: UNT Libraries Government Documents Department

Test of thick vessel with a flaw in residual stress field

Description: Intermediate test vessel V-8, a 152-mm-thick vessel fabricated of SA533, grade B, class 1 steel, was pressurized to failure at -23/sup 0/C. The vessel contained a fatigue-sharpened notch adjacent to a half-bead weld repair that had not been stress relieved. Residual stresses and fracture toughnesses were determined before the pressure test by measurements on a prototypical weld, and fracture predictions were made by linear elastic fracture analysis. Predictions agreed well with test results, demonstrating the important influence of high residual stresses on fracture behavior.
Date: January 1, 1979
Creator: Bryan, R.H.; Iskander, S.K.; Holz, P.P.; Merkle, J.G. & Whitman, G.D.
Partner: UNT Libraries Government Documents Department

An experimental study of the effect of stainless steel cladding on the structural integrity of flawed steel plates in bending

Description: A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Experimental results from tests on large clad and unclad plate specimens with surface flaws 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. The fracture surfaces of unclad plates suggest that the flaw evolves through alternately tunneling then breaking to the surface. In the case of clad plates, it is hypothesized that the tough, strong surface layer inhibits the tunneled flaw from propagating to the surface.
Date: January 1, 1989
Creator: Iskander, S.K.; Nanstad, R.K.; Robinson, G.C. & Oland, C.B.
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

Effects of 50/degree/C surveillance and test reactor irradiations on ferritic pressure vessel steel embrittlement

Description: The results of surveillance tests on the High-Flux Isotope Reactor (HFIR) pressure vessel at the Oak Ridge National Laboratory revealed that a greater than expected embrittlement had taken place after about 17.5 effective full-power years of operation and an operational assessment program was undertaken to fully evaluate the vessel condition and recommend conditions under which operation could be resumed. A research program was undertaken that included irradiating specimens in the Oak Ridge Research Reactor. Specimens of the A212 grade B vessel shell material were included, along with specimens from a nozzle qualification weld and a submerged-arc weld fabricated at ORNL to reproduce the vessel seam weld. The results of the surveillance program and the materials research program performed in support of the evaluation of the HFIR pressure vessel are presented and show the welds to be more radiation resistant than the A212B. Results of irradiated tensile and annealing experiments are described as well as a discussion of mechanisms which may be responsible for enhanced hardening at low damage rates. 20 refs., 22 figs., 5 tabs.
Date: January 1, 1988
Creator: Nanstad, R.K.; Iskander, S.K.; Rowcliffe, A.F.; Corwin, W.R. & Odette, G.R.
Partner: UNT Libraries Government Documents Department

Effects of irradiation on crack-arrest toughness of two high-copper welds

Description: The objective of this study is to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288{degree}C to an average fluence of 1.9 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV). A preliminary evaluation of the 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 seem to have been altered by irradiation. 10 refs., 9 figs., 7 tabs.
Date: January 1, 1990
Creator: Iskander, S.K.; Corwin, W.R. & Nanstad, R.K.
Partner: UNT Libraries Government Documents Department