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Pressurized-thermal-shock experiments: PTSE-1 results and PTSE-2 plans

Description: The first pressurized-thermal-shock experiment (PTSE-1) was performed with a vessel with a 1-m-long flaw in a plug of specially tempered steel having the composition of SA-508 forging steel. The second experiment (PTSE-2) will have a similar arrangement, but the material in which the flaw will be implanted is being prepared to have low tearing resistance. Special tempering of a 2 1/4 Cr - 1 Mo steel plate has been shown to induce a low Charpy impact energy in the upper-shelf temperature range. The purpose of PTSE-2 is to investigate the fracture behavior of low-upper-shelf material in a vessel under the combined loading of concurrent pressure and thermal shock. The primary objective of the experimental plan is to induce a rapidly propagating cleavage fracture under conditions that are likely to induce a ductile tearing instability at the time of arrest of the cleavage fracture. The secondary objective of the test is to extend the range of the investigation of warm prestressing. 11 figs.
Date: January 1, 1985
Creator: Bryan, R.H.; Nanstad, R.K.; Wanner, R.; Merkle, J.G.; Robinson, G.C. & Whitman, G.D.
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

Effects of irradiation temperature on Charpy and tensile properties of high-copper, low upper-shelf, submerged-arc welds

Description: This paper presents analyses of the Charpy impact and tensile test data, including adjustments for irradiation temperature and fluence normalization which make possible comparison of the irradiation sensitivity of the different welds. Analyses revealed dependence of yield and ultimate strength on irradiation temperature {minus}0.8 MPA/{degrees}C, respectively. Similarly, the Charpy impact energy changes due to irradiation temperature were {minus}0.5{degrees}C/{degrees}C for transition shift and {minus}0.05 J/{degrees}C for upper-shelf energy decrease. After adjustment to an irradiation temperature of 288{degrees}C and normalization to a fluence of 8 {times} 10{sup 18} neutrons/cm{sup 2} percentage increases in yield strength due to irradiation ranged from about 21 to 35% while those for ultimate strength ranged from about 13 to 20%. The Charpy transition temperature shifts ranged from 59 to 123{degrees}C while the postirradiation upper-shelf energies ranged from 58 to 79 J.
Date: December 31, 1992
Creator: Nanstad, R. K. & Berggren, R. G.
Partner: UNT Libraries Government Documents Department

The effect of aging at 343{degree}C on the mechanical properties and microstructure of type 308 stainless steel weldments

Description: The effect of long-term aging at intermediate temperatures on the mechanical properties of stainless steel welds has been studied. Three type 308 multipass shielded metal-arc welds with ferrite levels of 4, 8, and 12% were aged up to 20,000 h at 343C. Tensile tests showed little effect of aging on either the yield or ultimate tensile strengths, but the impact toughness was significantly degraded. The extent of the degradation increased with increasing ferrite content and increasing aging time. Examination of the microstructure with transmission electron microscopy and atom probe field-ion microscopy revealed that the ferrite phase had undergone spinodal decomposition as a result of aging. In addition, G-phase particles were observed at dislocations, and finer G-phase particles were homogeneously distributed throughout the ferrite phase. The changes in the mechanical properties and the fractography are discussed in light of the observed changes in the microstructure.
Date: December 31, 1992
Creator: Alexander, D. J.; Alexander, K. B.; Miller, M. K. & Nanstad, R. K.
Partner: UNT Libraries Government Documents Department

Effects of commercial cladding on the fracture behavior of pressure vessel steel plates

Description: The objective of this program is 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 thermal shock conditions. Preliminary results from testing at temperatures 10/sup 0/ and 60/sup 0/C below NDT have shown that (1) a tough surface layer (cladding and/or HAZ) 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.
Date: January 1, 1987
Creator: Iskander, S.K.; Alexander, D.J.; Bolt, S.E.; Cook, K.V.; Corwin, W.R.; Oland, B.C. et al.
Partner: UNT Libraries Government Documents Department

Preliminary results from Charpy impact testing of irradiated JPDR weld metal and commissioning of a facility for machining of irradiated materials

Description: Forty two full-size Charpy specimens were machined from eight trepans that originated from the Japan Power Demonstration Reactor (JPDR). They were also successfully tested and the preliminary results are presented in this report. The trends appear to be reasonable with respect to the location of the specimens with regards to whether they originated from the beltline or the core regions of the vessel, and also whether they were from the inside or outside regions of the vessel wall. A short synopsis regarding commissioning of the facility to machine irradiated materials is also provided.
Date: September 1, 1999
Creator: Iskander, S.K.; Hutton, J.T.; Creech, L.E.; Nanstad, R.K.; Manneschmidt, E.T.; Rosseel, T.M. et al.
Partner: UNT Libraries Government Documents Department

Use of precracked Charpy and smaller specimens to establish the master curve

Description: The current provisions used in the U.S. Code of Federal Regulations for the determination of the fracture toughness of reactor pressure vessel steels employs an assumption that there is a direct correlation between K{sub Ic} lower-bound toughness and the Charpy V-notch transition curve. Such correlations are subject to scatter from both approaches which weakens the reliability of fracture mechanics-based analyses. In this study, precracked Charpy and smaller size specimens are used in three-point static bend testing to develop fracture mechanics based K{sub k} values. The testing is performed under carefully controlled conditions such that the values can be used to predict the fracture toughness performance of large specimens. The concept of a universal transition curve (master curve) is applied. Data scatter that is characteristic of commercial grade steels and their weldments is handled by Weibull statistical modeling. The master curve is developed to describe the median K{sub Jc} fracture toughness for 1T size compact specimens. Size effects are modeled using weakest-link theory and are studied for different specimen geometries. It is shown that precracked Charpy specimens when tested within their confined validity limits follow the weakest-link size-adjustment trend and predict the fracture toughness of larger specimens. Specimens of smaller than Charpy sizes (5 mm thick) exhibit some disparities in results relative to weakest-link size adjustment prediction suggesting that application of such adjustment to very small specimens may have some limitations.
Date: December 1997
Creator: Sokolov, M. A.; McCabe, D. E.; Nanstad, R. K. & Davidov, Y. A.
Partner: UNT Libraries Government Documents Department

Thermal embrittlement of reactor vessel steels

Description: As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels.
Date: June 1, 1995
Creator: Corwin, W. R.; Nanstad, R. K.; Alexander, D. J.; Stoller, R. E.; Wang, J. A. & Odette, G. R.
Partner: UNT Libraries Government Documents Department

Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

Description: One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10{sup 19} n/cm{sup 2} (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10{sup 19} n/cm{sup 2} (>l MeV). In both cases, irradiations were conducted at {approximately}290 C and annealing treatments were conducted at {approximately}454 C. The ORNL and RRC-RI experiments have shown generally ...
Date: June 16, 1998
Creator: Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A. & Sokolov, M.A.
Partner: UNT Libraries Government Documents Department

Fracture-mechanics data deduced from thermal-shock and related experiments with LWR pressure-vessel material

Description: Pressurized water reactors (PWRs) are susceptible to certain types of hypothetical accidents that can subject the reactor pressure vessel to severe thermal shock, that is, a rapid cooling of the inner surface of the vessel wall. The thermal-shock loading, coupled with the radiation-induced reduction in the material fracture toughness, introduces the possibility of propagation of preexistent flaws and what at one time were regarded as somewhat unique fracture-oriented conditions. Several postulated reactor accidents have been analyzed to discover flaw behavior trends; seven intermediate-scale thermal-shock experiments with steel cylinders have been conducted; and corresponding materials characterization studies have been performed. Flaw behavior trends and related fracture-mechanics data deduced from these studies are discussed.
Date: January 1, 1982
Creator: Cheverton, R.D.; Canonico, D.A.; Iskander, S.K.; Bolt, S.E.; Holz, P.P.; Nanstad, R.K. et al.
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

Performance of low upper-shelf material under pressurized-thermal-shock loading (PTSE-2)

Description: The second pressurized-thermal-shock experiment (PTSE-2) of the Heavy-Section Steel Technology Program was conceived to investigate fracture behavior of steel with low ductile-tearing resistance. PTSE-2 was designed primarily to reveal the interaction of ductile and brittle modes of fracture and secondarily to investigate the effects of warm prestressing. A test vessel was prepared by inserting a crack-like flaw of well-defined geometry on the outside surface of the vessel. The flaw was 1 m long by approx.15 mm deep. The instrumented vessel was placed in a test facility in which it was initially heated to a uniform temperature and was then concurrently cooled on the outside and pressurized on the inside. These actions produced an evolution of temperature, toughness, and stress gradients relative to the prepared flaw that was appropriate to the planned objectives. The experiment was conducted in two separate transients, each one starting with the vessel nearly isothermal. The first transient induced a warm prestressed state, during which K/sub I/ first exceeded K/sub Ic/. This was followed by repressurization until a cleavage fracture propagated and arrested. The final transient was designed to produce and investigate a cleavage crack propagation followed by unstable tearing. During this transient the fracture events occurred as had been planned. 7 refs., 13 figs., 2 tabs.
Date: January 1, 1987
Creator: Bryan, R.H.; Bass, B.R.; Bolt, S.E.; Bryson, J.W.; Corwin, W.R.; Nanstad, R.K. et al.
Partner: UNT Libraries Government Documents Department

Chemical composition and RT[sub NDT] determinations for Midland weld WF-70

Description: The Heavy-Section Steal Irradiation Program Tenth Irradiation Series has the objective to investigate the affects of radiation on the fracture toughness of the low-upper-shelf submerged-arc welds (B W designation WF-70) in the reactor pressure vessel of the canceled Midland Unit 1 nuclear plant. This report discusses determination of variations in chemical composition And reference temperature (RT[sub NDT]) throughout the welds. Specimens were machined from different sections and through thickness locations in both the beltline and nozzle course welds. The nil-ductility transition temperatures ranged from [minus]40 to [minus]60[degrees]C ([minus]40 and [minus]76[degrees]F) while the RT[sub NDT]S, controlled by the Charpy behavior, varied from [minus]20 to 37[degrees]C ([minus]4 to 99[degrees]F). The upper-shelf energies varied from 77 to 108 J (57 to 80 ft-lb). The combined data revealed a mean 41-J (30-ft-lb) temperature of [minus]8[degrees]C (17[degrees]F) with a mean upper-shelf energy of 88 J (65 ft-lb). The copper contents range from 0.21 to 0.34 wt % in the beltline weld and from 0.37 to 0.46 wt % in the nozzle course weld. Atom probe field ion microscope analyses indicated substantial depletion of copper in the matrix but no evidence of copper clustering. Statistical analyses of the Charpy and chemical composition results as well as interpretation of the ASME procedures for RT[sub NDT] determination are discussed.
Date: December 1, 1992
Creator: Nanstad, R.K.; McCabe, D.E.; Swain, R.L. & Miller, M.K. (Oak Ridge National Lab., TN (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

Use of forces from instrumented Charpy V-notch testing to determine crack-arrest toughness

Description: The objective of this investigation is an estimation of the crack-arrest toughness, particularly of irradiated materials, from voltage versus time output of an instrumented setup during a test on a Charpy V-notch (CVN) specimen. This voltage versus time trace (which can be converted to force versus displacement) displays events during fracture of the specimen. Various stages of the fracture process can be identified on the trace, including an arrest point indicating arrest of brittle fracture. The force at arrest, F{sub a}, versus test temperature, T, relationship is examined to explore possible relationships to other experimental measures of crack-arrest toughness such as the drop-weight nil-ductility temperature (NDT), or crack-arrest toughness, K{sub a}. For a wide range of weld and plate materials, the temperature at which F{sub a} = 2.45 kN correlates with NDT with a standard deviation, sigma, of about 11 K. Excluding the so-called low upper-shelf energy (USE) welds from the analysis resulted in F{sub a} = 4.12 kN and {sigma} = 6.6 K. The estimates of the correlation of the temperature for F{sub a} = 7.4 kN with the temperature at 100-MPa{radical}m level for a mean American Society of Mechanical Engineers (ASME) type K{sub Ia} curve through crack-arrest toughness values show that prediction of conservative values of K{sub a} are possible.
Date: June 1996
Creator: Iskander, S. K.; Nanstad, R. K.; Sokolov, M. A.; McCabe, D. E. & Hutton, J. T.
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

Investigation of the bases for use of the K sub Ic curve

Description: Title 10 of the Code of Federal Regulations, Part 50 (10CFR50), Appendix G, establishes the bases for setting allowable pressure and temperature limits on reactors during heatup and cooldown operation. Both the K{sub Ic} and K{sub Ia} curves are utilized in prescribed ways to maintain reactor vessel structural integrity in the presence of an assumed or actual flaw and operating stresses. Currently, the code uses the K{sub Ia} curve, normalized to the RT{sub NDT}, to represent the fracture toughness trend for unirradiated and irradiated pressure vessel steels. Although this is clearly a conservative policy, it has been suggested that the K{sub Ic} curve is the more appropriate for application to a non-accident operating condition. A number of uncertainties have been identified, however, that might convert normal operating transients into a dynamic loading situation. Those include the introduction of running cracks from local brittle zones, crack pop-ins, reduced toughness from arrested cleavage cracks, description of the K{sub Ic} curve for irradiated materials, and other related unresolved issues relative to elastic-plastic fracture mechanics. Some observations and conclusions can be made regarding various aspects of those uncertainties and they are discussed in this paper. A discussion of further work required and under way to address the remaining uncertainties is also presented.
Date: January 1, 1991
Creator: McCabe, D.E.; Nanstad, R.K. (Oak Ridge National Lab., TN (USA)); Rosenfield, A.R.; Marschall, C.W. (Battelle, Columbus, OH (USA)) & Irwin, G.R. (Maryland Univ., College Park, MD (USA))
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Materials Research and Development Program Plan, Revision 4

Description: DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 950°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Some of the general and administrative aspects of the R&D Plan include: • Expand American Society of Mechanical Engineers (ASME) Codes and American Society for Testing and Materials (ASTM) Standards in support of the NGNP Materials R&D Program. • Define and develop inspection needs and the procedures for those inspections. • Support selected university materials related R&D activities that would be of direct benefit to the NGNP Project. • Support international materials related collaboration activities through the DOE sponsored Generation IV International Forum (GIF) Materials and Components (M&C) Project Management Board (PMB). • Support document review activities through the Materials Review Committee (MRC) or other suitable forum.
Date: September 1, 2007
Creator: Hayner, G. O.; Bratton, R. L.; Mizia, R. E.; Windes, W. E.; Corwin, W. R.; Burchell, T. D. et al.
Partner: UNT Libraries Government Documents Department