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Heavy Section Steel Technology Program. Part I. Program scope, fracture, materials and thermal shock studies

Description: A brief summary of the scope of HSST projects is presented. A major emphasis throughout the course of this work is the verification of methods of fracture prediction which can be utilized in an assessment of pressure vessel integrity. The utility of small specimens which are appropriate for surveillance purposes continues to be investigated in both the unirradiated and post- irradiated conditions. 11 references (auth)
Date: January 1, 1975
Creator: Whitman, G.D.
Partner: UNT Libraries Government Documents Department

Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

Description: The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references (auth)
Date: January 1, 1975
Creator: Whitman, G.D.
Partner: UNT Libraries Government Documents Department

Pressurized-thermal-shock tests

Description: Pressurized-thermal-shock experiments are required to validate methods of fracture analysis to establish the degree of conservatism or accuracy involved in predictions of flaw behavior under certain accident conditions. By using methods and facilities developed for this purpose we can simulate materials and loading regimes to evaluate the integrity of flawed reactor pressure vessels subjected to pressurized-thermal-shock transients. These accidents involve small-break loss-of-coolant accidents, steamline breaks, and other similar overcooling accident scenarios involving combined temperature and pressure transients.
Date: January 1, 1981
Creator: Whitman, G.D.
Partner: UNT Libraries Government Documents Department

Pressurized-thermal-shock experiments. [PWR]

Description: The primary objective of the ORNL pressurized-thermal-shock (PTS) experiments is to verify analytical methods that are used to predict the behavior of pressurized-water-reactor vessels under these accident conditions involving combined pressure and thermal loading. The criteria on which the experiments are based are: scale large enough to attain effective flaw border triaxial restraint and a temperature range sufficiently broad to produce a progression from frangible to ductile behavior through the wall at a given time; use of materials that can be completely characterized for analysis; stress states comparable to the actual vessel in zones of potential flaw extension; range of behavior to include cleavage initiation and arrest, cleavage initiation and arrest on the upper shelf, arrest in a high K/sub I/ gradient, warm prestressing, and entirely ductile behavior; long and short flaws with and without stainless steel cladding; and control of loads to prevent vessel burst, except as desired. A PTS test facility is under construction which will enable the establishment and control of wall temperature, cooling rate, and pressure on an intermediate test vessel (ITV) in order to simulate stress states representative of an actual reactor pressure vessel.
Date: January 1, 1982
Creator: Whitman, G.D. & McCulloch, R.W.
Partner: UNT Libraries Government Documents Department

Heavy-Section Steel Technology Program quarterly progress report, January-March 1980

Description: The program comprises studies related to all areas of the technology of materials fabricated into thick-section primary-coolant containment systems of light-water-cooled nuclear power reactors. The principal area of investigation is the behavior and structural integrity of steel pressure vessels containing cracklike flaws. Current work is organized into the following tasks: (1) program administration and procurement, (2) fracture mechanics analyses and investigations, (3) investigations of irradiated materials, (4) thermal shock investigations, and (5) pressure vessel investigations. Work performed under the existing research and development subcontracts is included in this report.
Date: July 1, 1980
Creator: Whitman, G.D. & Bryan, R.H.
Partner: UNT Libraries Government Documents Department

Heavy-Section Steel Technology Program quarterly progress report for April-June 1980

Description: The Heavy-Section Steel Technology Program is an engineering research activity conducted by the Oak Ridge National Laboratory for the Nuclear Regulatory Commission. The program comprises studies related to all areas of the technology of materials fabricated into thick-section primary-coolant containment systems of light-water-cooled nuclear power reactors. The investigation focuses on the behavior and structural integrity of steel pressure vessels containing cracklike flaws. Current work is organized into five tasks: (1) program administration and procurement, (2) fracture mechanics analyses and investigations, (3) investigations of irradiated materials, (4) thermal shock investigations, and (5) pressure vessel investigations. Nozzle-corner cracks under combined pressure and thermal loadings are being analyzed. Mechanisms of damping in crack propagation are being studied. Irradiation of the first specimens in the Fourth HSST Irradiation Series continued, and impact tests of several Charpy specimens from the previous series were completed. Heat-treatment conditions for the next thermal shock test were selected, and preparation of the test cylinder was initiated. Work was initiated to develop a low-upper-shelf seam weld for intermediate test vessel V-8A, and facility planning for pressurized thermal shock tests continued. 16 refs., 29 figs., 11 tabs.
Date: October 1, 1980
Creator: Whitman, G.D. & Bryan, R.H.
Partner: UNT Libraries Government Documents Department

Evaluations of half-bead weld repair procedures with thick-wall pressure vessels

Description: The results of research on the evaluation of the half-bead weld repair method for use on nuclear reactor components are reviewed from data obtained on thick-section test pieces and intermediate-size pressure vessels. Material properties, the magnitude of residual stresses and the structural behavior of flawed pressure vessels are being obtained to determine the adequacy of the weld repair method for application in thick-section components.
Date: January 1, 1978
Creator: Canonico, D.A. & Whitman, G.D.
Partner: UNT Libraries Government Documents Department

HSST crack-arrest studies overview

Description: An overview is given of the efforts underway in the Heavy-Section Steel Technology (HSST) Program to better understand and model crack-arrest behavior in reactor pressure vessel steels. The efforts are both experimental and analytical. The experimental work provides K/sub Ia/ data from laboratory-sized specimens, from thick-wall cylinders which exhibit essentially-full restraint and from nonisothermal wide-plate specimens. These data serve to define toughness-temperature trends and to provide validation data under prototypical reactor conditions. The analytical efforts interpret and correlate the data, plus provide LEFM, elastodynamic and viscoplastic methods for analyzing crack run-arrest behavior in reactor vessels. The analysis methods are incorporated into finite element computer programs which are under development at three separate laboratories. 22 refs., 10 figs.
Date: January 1, 1985
Creator: Pugh, C.E. & Whitman, G.D.
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

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

Test of 6-inch-thick pressure vessels. Series 1: intermediate test vessels V-1 and V-2

Description: The intermediate vessel tests have been subdivided into four seriesi flaws in cylindrical vessels, A508, class 2 forging steel-two vessels; flaws in cylindrical vessels with longitudinal weld seams, A508, class 2 forging steel, submerged-arc welds-three vessels; flaws in cylindrical vessels wlth longitudinal weld seams, A533, grade B, class l plate steel, submerged-arc weld-two vessels; and cylindrical vessels with radially attached nozzles, vessels of A508, chass 2 forging steel and A533, grade B, class 1 plate steel; nozzle of A508 class 2 forging steel-three vessels. A comprehensive description of the pertinent factors considered in the design of the vessels is presented. Construction of the test facility and documentation of test results and fracture predictions are included. Emphasis is placed on providing the test results in such a manner that they form a resource for amy investigators interested in the problem of fracture. (auth)
Date: February 1, 1974
Creator: Derby, R.W.; Merkle, J.G.; Robinson, G.C.; Whitman, G.D. & Witt, F.J.
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

Pressurized-thermal-shock experiments with thick vessels

Description: Information is provided on the series of pressurized-thermal-shock experiments at the Oak Ridge National Laboratory, motivated by a concern for the behavior of flaws in reactor pressure vessels having welds or shells exhibiting low upper-shelf Charpy impact energies, approx. 68J or less. (JDB)
Date: January 1, 1986
Creator: Bryan, R.H.; Nanstad, R.K.; Merkle, J.G.; Robinson, G.C. & Whitman, G.D.
Partner: UNT Libraries Government Documents Department

HSST pressurized-thermal-shock experiment, PTSE-1. [PWR; BWR]

Description: The first pressurized-thermal-shock experiment (PTSE-1) in the Heavy-Section Steel Technology (HSST) Program is the most recent of a long successtion of fracture-mechanics experiments that are on a scale that allows important aspects of fracture behavior of reactor pressure vessels to be simulated. Such experiments are the means by which theoretical models of fracture behavior can be evaluated for possible aplication to fracture analysis of vessels in nuclear plants. The principal issues of concern in the pressurized-thermal-shock experiments are: (1) warm prestressing phenomena, (2) crack propagation from brittle to ductile regions, (3) transient crack stabilization in ductile regions, and (4) crack shape changes in bimetallic zones of clad vessels. PTSE-1 was designed to investigate the first three issues under conditions relevant to a flawed reactor vessel during an overcooling accident.
Date: January 1, 1984
Creator: Bryan, R.H.; Bass, B.R.; Robinson, G.C.; Merkle, J.G.; Whitman, G.D. & Pugh, C.E.
Partner: UNT Libraries Government Documents Department

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

Test of 6-in. -thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading. [BWR; PWR]

Description: HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7.
Date: February 1, 1978
Creator: Bryan, R.H.; Cate, T.M.; Holz, P.P.; King, T.A.; Merkle, J.G.; Robinson, G.C. et al.
Partner: UNT Libraries Government Documents Department

Results and conclusions from the first pressurized-thermal-shock experiment. [Overcooling]

Description: Pretest estimates of fracture toughness are reasonably close to the PTSE-1 values. Furthermore, the ASME Sect. XI toughness relationships are conservative relative to actual material characteristics. The experiment demonstrated that arrest toughness substantially above the 220 MPa...sqrt..m cutoff of Sect. XI could be realized. The arrest values in PTSE-1 also are consistent with arrest measurements made in wide-plate tests and reported by the Japan Welding Council. The highest PTSE-1 value of arrest occurred at a temperature approx. 30 K above the onset of the Charpy upper shelf. This is believed to be very close to the threshold temperature above which cleavage fracture cannot persist. This result also suggests that the methods of linear elastic fracture mechanics have an important role in fracture evaluation at high (upper-shelf) temperatures. The PTSE-1A and -1B transients were a demonstration that simple warm prestressing (K/sub I/ < 0) strongly inhibits crack initiation. With allowance for uncertainty in the true K/sub Ic/ values it is evident that K/sub I/ exceeds K/sub Ic/ during warm prestressing by 50% to 90%. Thus, the effectiveness of simple warm prestressing has now been demonstrated in two experiments with thick cylinders, thermal shock experiment TSE-5A and PTSE-1. In the A transient, simple anti-warm prestressing (K/sub I/ > 0) prevailed during two periods of 40-s and 60-s duration without crack initiation, although K/sub I/ exceeded K/sub Ic/ by 30% to 50%. Clearly simple anti-warm prestressing is not a sufficient condition to alleviate the effects of warm prestressing. A narrow band of ductile tearing formed ahead of the initial cleavage fracture. The conclusions drawn from PTSE-1 suggest that procedures used for evaluating overcooling accidents in pressurized-water reactors should take into consideration realistically the fracture mechanisms that have been clearly demonstrated but not yet generally accepted.
Date: January 1, 1984
Creator: Bryan, R.H.; Bolt, S.E.; Merkle, J.G.; Bass, B.R.; Bryson, J.W.; Robinson, G.C. et al.
Partner: UNT Libraries Government Documents Department