The USAEC heavy section steel technology program: Objectives and status

NUCLEAR ENGINEERING AND DESIGN 20 (1972) 169-180. NORTH-HOLLAND PUBLISHING COMPANY

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The Heavy Section Steel Technology (HSST) program is concerned with the primary vessels for water reactors with the main thrust being placed on the effects of flaws, material inhomogeneities and discontinuities on the behavior of the vessels under normal operating and accident conditions. Emphasis is placed on developing a technology that can be used to establish the margin of safety against failure of the vessels, particularly catastrophic vessel failure beyond the control of plant safety features. Fracture phenomena under the environmental conditions of interest and the effects of section thickness on the fracture behavior of the large reactor vessels are being examined through a systematic progression of specimen sizes up to the equivalent of a full section thickness of 12 in. The program will culminate in a series of simulated service tests in which deliberately flawed vessels of substantial thickness will be tested to fracture. In this paper the status of the various investigations will be presented and the significant accomplishments in each of the areas of investigation will be summarized.

1. Introduction With the emergence of nuclear reactors as one o f the major heat resources, the United States electrical power industry has been quick to recognize the potential of nuclear reactor systems for power generation. Prior to 1971, the peak period for nuclear orders was in 1967 when two-thirds of the announced power generating plants were nuclear. It is now conservatively estimated that by 1980 about 25% of the total electrical output in the United States will be from nuclear plants. At present in the United States over 100 civilian nuclear power stations are in operation (approximately 20), under construction (approximately 55), and on order (approximately 35). The total electrical capacity in service exceeds 8300 MWe. Some of

* Research performed for the Oak Ridge National Laboratory, which is operated by Union Carbide Corporation for the U. S. Atomic Energy Commission.

the larger plants now under contruction have capacities exceeding 1100 MWe. The two major systems which have evolved to commercial status in the United States are the boiling water reactor (BWR) system and the pressurized water reactor (PWR) system. Concurrent with the growth of the nuclear power industry has been the increase in size of the plants themselves. This increase has resulted mainly from scaling up small systems with some modification in pressures and operating temperatures. As a consequence the pressure vessels o f the systems have increased in diameter and thickness. The thickness o f the plate for fabricating some larger pressurized water vessels approaches 12 in with diameters of the vessels from 14 to 16 ft. The standard operating temperature is around 550 ° F. Pressures range upward to around 2500 psi for the PWR systems, whereas for BWR systems the vessels are designed for about half that pressure. Equally significant to the increased

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FJ. Witt, The USAECheavy section steel technology program

sizes is the economic incentive to place these large plants near densely populated and industrial areas. The main concern of the heavy section steel technology (HSST) program is with the massive pressure vessels with emphasis being placed on the effects of flaws, material inhomogeneities and discontinuities on the behavior of the vessels under startup, operating, cooldown and accident conditions. This interest stems from the fact that, inherently, flaws (discontinuities) exist in vessels though possibly quite small and that in-service inspection techniques require further development if applicability is desired for the service life of the plants, some being designed to operate 40 years. It is the objective of the HSST program to develop the technology necessary to establish a reliable means for estimating a conservative margin of safety against fracture for nuclear pressure vessels during the service life of the plants, particularly such fracture as might endanger the public. Research and development under the HSST program is of necessity broad based because of the extensive interrelating fields of study which are involved; that is, metallurgy, chemistry, material properties, inspection, analytical and experimental stress and strain analyses, environmental effects, fracture mechanics and the general area of fracture behavior. The material used to fabricate most American water reactor vessels is the ASTM A 533 grade B, class 1 plate steel or the equivalent forging grade steel ASTM A 508, class 2 [ 1]. The welds most commonly used are the submerged arc or the shielded metal arc weld. Electroslag welds have been used in some vessels. All these steels and weldments are being investigated but the major emphases are on the plate steel and the sub. merged arc welds. The HSST program is coordinated with efforts by other government agencies and the manufacturing and utility sectors of the nuclear power industry both in the United States and abroad. These total efforts were recently summarized in issues of Nuclear Engineering and Design [2,3] and should culminate in the quantification of safety assessments which are needed by the USAEC regulatory bodies, the professional code-writing bodies and the nuclear power industry. The activities of the HSST program are carried out under twelve separate tasks. About three-fourths of the task activities are performed under subcontract by research facilities throughout the United States.

A panel of twenty-eight scientists and engineers with expertise throughout the gamut of material technology has been organized to serve in an advisory capacity to the HSST program. Detailed discussions of HSST program investigations may be found in HSST semiannual progress reports [4], technical and documentary reports [5-22], and technica! manuscripts [23-35]. The most current technical publications (a collection of 13 papers) may be found in a recent issue of Nuclear Engineering and Design [36-48]. Information meetings are held on an annual basis; the abstracts of the papers presented at the 1971 meeting are given in ref. [3]. It is the intent of this paper to summarize the more pertinent results obtained under the auspices of the HSST program leaving the details to be found in the references given. Special emphasis will be placed on the results of very current research in an effort to supplement and summarize the more available data.

2. Material investigations The ASTM A 533, grade B, class 1 plate material together with welds was selected to be extensively investigated first under the HSST program. Consequently four plates approximately 10 .ft wide and 20 - ft long by 12-in thick were.purchased along with some 50 linear ft of weldment. The heat treatments, inspections and properties of these materials have been well documented [20-22]. Essentially the plates were found to be well within the specifications with yield strength at room temperatures approaching 70 000 psi and with ultimate strength between 85 000 and 90 000 psi. Extensive tensile, drop weight, nil-ductility and Charpy impact data have been obtained on the plates and welds [28,37]. A comparison of impact data among the various product forms is shown in fig. 1. It is seen that the interior of the plate has the highest transition temperature but reaches a high Charpy impact shelf level at around 150 ° F (65° C). In addition to the Charpy impact tests, dynamic tear tests and fracture toughness KIc and Kid tests have been performed on specimens up to 12-in thick [10,11,17,38]. Impact tests on nil.ductility transition temperature drop weight specimens scaled up to 4-in thick have also been performed [6]. The size of

F.Z Witt, The USAEC heavy section steel technology program 160

for irradiation [53]. A partial,summary of results is shown in fig. 5. Here it is seen that the reactor environments did not have adverse effects on the fatigue crack growth of A 533, grade B, class 1 steel plate or weldments.

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A primary objective of the HSST program is to develop methods for predicting fracture throughout the behavioral regimes of the materials; that is, for frangible, transitional and tough behavior. Linear elastic fracture mechanics is applicable to the frangible behavior; however, it is necessary that stress intensity factors be obtained for the configurations of interest. Such factors are being obtained by testing epoxy models of the pertinent geometry. Both two-dimensional [8] and three-dimensional [16, 41] elastic-plastic finite-element computer analyses have been developed. The application of these analyses to fracture behavior is currently under study. The gross strain method of fracture analysis is also being studied [7, 18,40]. The majority of the work has been on flawed tensile type specimens but current work emphasizes combinations of tensile and bending stresses. The equivalent energy method has been developed for fracture analysis [54]. This procedure has been successfully used to predict the behavior of a series of large flawed tensile specimens. The specimen and fracture surfaces are shown in fig. 6. With the exception of specimen no. 4 which behaved in a much tougher manner than an almost similar specimen (no. 7), the energy to maximum load to fracture was successfully calculated [55]. These results are futher discussed in refs. [43] and [56]. Series of 1-in, 4-in, 6-in, and 9-in thick compact tension specimens have been tested for temperatures producing frangible behavior (temperatures at which fracture toughness KIc were obtained) as well as in the tough range up to 550 ° F, the operating temperature of water reactors [52]. Fracture surfaces from two of these specimens are seen in fig. 7. As for the flawed tensile specimens discussed above, the significance of the results to fracture analyses is discussed in ref. [56]. In both cases, however, the increase in toughness from fiat fracture at around 100 ° F (38 ° C) is vividly demonstrated.

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4. Simulated service tests The investigations summarized in this paper are hardware oriented in the sense that results and conclusions are meant to be applied to specific structures, water reactor pressure vessels. To this end the simulated service test portion of the HSST program is directed. The validation of the research is proceeding in steps from conventional, although perhaps thick, specimens to pressure vessels, with an intermediate series of large 6-in thick flawed tensile specimens, some fracture surfaces of which are seen in fig. 6. Complementing these large tensile tests are series of tests of 1/6 -scale model specimens as shown in fig. 8. A specimen before and one after fracture at 50 ° F are shown. The energy ratios for these two series of tests have been extensively studied, referenced to similar behavior of compact tension specimens (see ref. [56] ). As mentioned previously, the results of the tensile tests were, in general, successfully predicted. One of the major testing difficulties encountered was that of introducing sharp flaws in the tensile

specimens and pressure vessels, both small and large. A procedure for fatigue sharpening a crack by local pressurization was successfully developed for the larger specimens [57]. A metallurgical procedure consisting of electrically inducing hydrogen embrittlement in an electron beam weld has been used to produce cracks in the smaller specimens [58]. A series of 6-in thick pressure vessels with 39-in outside diameter, some with and some without nozzles, will be tested. A picture of one of the vessels is seen in fig. 9. Small models of these vessels are also being tested as reported in refs. [59] and [60]. Two 1-in thick vessels are seen in fig. 8, one before and one after fracture at 50 ° F. It is expected that from these tests the fracture behavior of the large vessels will be successfully predicted. Since they are fabricated from A 508 class 2 forging grade steel, predicting the behavior of these vessels will be an excellent test of the technology developed on plate material. The data pertinent to the forging steel fracture behavior will of course have to be obtained. The testing of these vessels is scheduled for initiation in early 1972.

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Fig. 6. Fracture surfaces of ten flawed tensile specimens (18-in wide by 6-in thick) from center material of A 533 grade B class 1 plate tested at temperatures between 50 ° F (no. 3) and 220° F (nos. 1,2, and 9). Specimen nos. 1 through 7 were longitudinally oriented while nos. 8 through 10 were transversely oriented. Generalizations to behaviors of highly irradiated vessels are also anticipated. With the successful completion of these tests, the major objectives of developing quantitative methods for assessing margins of safety under design and abnormal operating conditions should be firmly established.

5. Summary Extensive studies of the structural integrity of water reactor pressure vessels have been undertaken by the HSST program. Both transition temperature and brittle fracture approaches to safety analyses and fracture prevention are being synthesized into a quantitative understanding of the parameters which govern fracture. This understanding has resulted in a rapidly developing technology for assessing adequate margins of safety for reactor vessels under in-service conditions.

One of the major causes for concern in assessing the safety of large thick-walled reactor vessels is the uncertainty of the effects of size and thickness on fracture behavior. The research approach of testing series of a relatively diverse range of specimen configurations, sizes, and loading methods shows good promise for setting up lines of parallel fracture behavior when plotted against a size parameter. It also has been shown that this parallelism exists when testing temperature is made the controlling variable. The results so far obtained show good possibilities that typical pertinent trends of fracture behavior are consistent and can be extended into regions that cannot be tested full size within the practical limits of time, size, cost, temperature control, fluence level, and/or capacity of testing facilities. Investigations of the pertinent plates and weldments on more or less standard specimens, although perhaps

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Fig. 9. A 6-in thick pressure vessel near completion.

large, have clearly demonstrated that high toughness levels are reached at moderately low temperatures (200 ° F, 93 ° C and below) by the material even in 12-in ticknesses if the product form has not been irradiated. Irradiation is known to reduce the toughness; however, present indications are that sufficiently high toughness levels will be shown to exist at temperatures compatible with reactor operating requirements even for high irradiation fluence levels. More extensive investigations of the effect of irradiation are currently in progress. Methods of fracture analysis are being quantitized by comparison with results from the flawed tensile tests and the small flawed vessel tests. The generality of the methods developed are to be validated by the 6-in thick vessel tests which are to be initiated in 1972. With the successful completion of this phase of the program together with obtaining the pertinent irradiation data, it is expected that quantitative procedures will be developed by code-writing and regulatory bodies for guaranteeing both the safe and economical operation of water reactor nuclear power plants.

References 11] 1971 Annual Book of ASTMStandards, Part 4, American Society for Testing and Materials, (April 1971) 594,646. [ 2] Information, Abstracts of papers presented at the Fourth Semiannual Information Meeting of the HeavySection Steel TechnologyProgram, Nuclear Engineering and Design, 14, 2 (1970) 352. [ 3] information, Abstracts of papers presented at the Fifth Annual Information Meeting of the Heavy Section Steel Technology Program, Nuclear Engineering and Design, 17, I (1971) 170. [4] F.J.Witt, Heaw/Section Steel Technology Program Semiannual Progress Report, Aug. 31, 1967, USAEC Report ORNL-4176; Feb. 29, 1968, USAEC Report ORNL-4315; Aug. 31, 1968, USAEC Report ORNL-4377; Feb. 28, 1969, USAEC Report ORNL-4463;Aug. 31, 1969, USAECRe~ port ORNL-4512;Feb. 28, 1970, USAECReport ORNL4590; Aug. 31, 1970, USAECReport ORNL-4653,Oak Ridge National Laboratory. [5] S. Yukawa, Evaluation of Periodic Proof Testing and Warm Prestressing Procedures for Nuclear Reactor Vessels, Report HSSTP-TR-1,General Electric Company, Schenectady, New York, July 1, 1969. 16l L.W. Loechel,The Effect of Section Size on the Transition Temperature in Steel, Report No. MCA-69-189, Maxtin Marietta Company, Denver, Colorado, 1969.

F.J. ICitt, The USAEC heavy section steel technology program [7] P.N. Randall, Gross Strain Measure of Fracture Toughness of Steels, Report No. HSSTP-TR-3,TRW Systems Groups, Redondo Beach, California, November 1, 1969. [8] C. Visser, S.E. Gabrielse, and W. Van Buren, A TwoDimensional Elastic-Plastic Analysis of Fracture Test Specimens, Report No. WCAP-7368,Westinghouse Electric Corporation, PWR Systems Division, Pittsburgh, Pennsylvania, 1969. [9] T.R. Mager, F.O. Thomas and W.S. Hazelton, Evaluation by Linear Elastic Fracture Mechanics of Radiation Damage to Pressure Vessel Steels, Report No. WCAP7328, Revised, Westinghouse Electric Corporation, PWR Systems Division, Pittsburgh, Pennsylvania, October 1969. [10] W.O. Shabbits, W.H. Pryle and E.T. Wessel, Heavy Section Fracture Toughness Properties of A 533 Grade B, Class 1 Steel Plate and Submerged Axe Weldment, Report No. WCAP-7414,Westinghouse Electric Corporation, PWR Systems Division, Pittsburgh, Pennsylvania, Deoember 1969. [ 11 ] F.J. Loss, Dynamic Tear Test Investigations of the Fracture Toughness of Thick-Section Steel, NRL Report 7056 (also HSST-TR-7), U.S. Naval Research Laboratory. [12] P.B. Crosley and E.J. Ripling, Crack Arrest Fracture Toughness of A 533 Grade B Class 1 Pressure Vessel Steel, Report No. HSSTP-TR-8, Materials Research Laboratory, March 1970. [13] T.R. Mager, Post Irradiation Testing of 2 T Compact Tension Specimens, WCAP-7561, Westinghouse Electric Corporation, PWR Systems Division, Pittsburgh, Pennsylvania, August 1970. [14] T.R. Mager, Fracture Toughness Characterization Study of A 533 Class 1 Steel, WCAP-7558,Westinghouse Electric Corporation, PWR Systems Division, Pittsburgh, Pennsylvania, September 1970. [151 T.R. Mager, Notch Preparation in Compact Tension Specimens, WCAP-7579,Westinghouse Electric Corporation, PWR Systems Division, Pittsburgh, Pennsylvania, November 1970. [16] N. Levy and P.V. Marcal, Three-Dimensional ElasticPlastic Stress and Strain Analysis for Fracture Mechanics, Phase 1: Simple Flawed Specimens, HSSTP-TR-12, Brown University, Providence, Rhode Island, December 1970. [ 17 ] W.O. Shabbits, Dynamic Fracture Toughness Properties of Heavy Section A 533 Grade B Class 1 Steel Plate, WCAP-6723, Westinghouse Electric Corporation, PWR Systems Division, Pittsburgh, Pennsylvania (December 1970). [18] P.N. Randall, Gross Strain Crack Tolerance of A.533-.B Steel, Report No. HSSTP-TR-14,TRW Systems Group, Redondo Beach, California, May 1, 1971. [19] H.T. Corten and R.H. Sailors, Relationship Between Material Fracture Toughness Using Fracture Mechanics and Transition Temperature Tests, T. & A.M. Report No. 346, University of Illinois, Urbana, Illinois (August 1, 1971).

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[ 20] C.E. Childress, Fabrication History of the First Two 12-in.-Thick ASTM A 533, Grade B, Class 1 Steel Plates of the Heavy Section Steel Technology Program, Documentary Report 1, USAEC Report ORNL-4313, Oak Ridge National Laboratory, Febrgary 1969. [21 ] C.E. Childress, Fabrication History of the Third and Fourth 12-in.-Thick ASTM A 533, Grade B, Class 1 Steel Plates of the Heavy Section Steel Technology Program, Documentary Report 2, USAEC Report ORNL-4313, Pt. 2, Oak Ridge National Laboratory, February 1970. [22] C.E. Childress, Fabrication Procedures and Acceptance Data for ASTM A 533 Welds and a 10-in.-Thick ASTM A 543 Plate of the Heavy Section Steel Technology Program, Documentary Report 3, USAEC Report ORNL-4313, Pt. 3, Oak Ridge National Laboratory, January 1971. [23] HSST Intermediate Vessel Closure Analysis, Technical Report E-1253 (b), Teledyne Materials Research Company, March 25, 1970. [24] C.L. Segaser, Conceptual System Design Description of the Intermediate Vessel Tests for the Heavy Section Steel Technology Program, USAEC Report ORNL-TM2849, Oak Ridge National Laboratory, June 1970. [25] F.J. Witt and R.G. Berggren, Siz~ Effects and Energy Disposition in Impact Specimen Testing of ASTM A 533, Grade B Steel, USAEC Report ORNL-TM-3030, Oak Ridge National Laboratory, August 1970. [26] D.A. Canonico, Transition Temperature Considerations for Thick-WallNuclear Pressure Vessels, USAEC Report ORNL-TM-3114, Oak Ridge National Laboratory, October 1970. [271 G.D. Whitman and F.J. Witt, Heavy Section Steel Technology Program, USAEC Report ORNL-TM3055, Oak Ridge National Laboratory, October 1970. [28] D.A. Canonico and R.G. Berggren, Tensile and Impact Properties of Thick Section Plate and Weldments, USAEC Report ORNL-TM-321l, Oak Ridge National Laboratory, January 1971. [29] C.E. Childress, Manual for ASTM A 533 Grade B Class 1 Steel (HSST Plate 03) provided to the International Atomic Energy Agency, USAEC Report ORNL-TM-3193, Oak Ridge National Laboratory, March 1971. [30] C.L. Segaser, Feasibility Study - Irradiation of Heavy Section Steel Specimens in the South Test Facility of the Oak Ridge Research Reactor, USAEC Report ORNL-TM-3234, Oak Ridge National Laboratory, May 1971. [31] R.W. Derby and C.L. Segaser, Quality Assurance Program Plan, Intermediate Vessel Test Facility, USAEC Report No. ORNL-TM-3373, Oak Ridge National Laboratory, May 1971. [32] G.C. Robinson, Discussion of SwRI Model Parametric Tests, USAEC Report No. ORNL-TM-3313, Oak Ridge National Laboratory, June 1971.

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F.J. Witt, The USAEC heavy section steel technology program

[ 33] C.W. Hunter and J.A. Williams, Fracture and Tensile Behavior of Neutron-Irradiated A 533-B Pressure Vessel Steel, USAEC Report HEDL-TME 71-76, Hanford Engineering Development Laboratory, February 5, 1971. [34] J.G. Merkle, L.F. Kooistra and R.W. Derby, Interpretation of the Drop-Weight Test in Terms of Strain Tolerance (Gross Strain) and Fracture Mechanics, USAEC ORNLTM-3247, Oak Ridge National Laboratory, June 1971. [35] N. Krishnamurthy, Three-Dimensional Finite Element Analysis of Thick-Walled Vessel-Nozzle Junctions with Curved Transition, USAEC Report ORNL-TM-3315, Oak Ridge National Laboratory, July 1971. [36] F.J. Witt, Introduction: HSST Program Investigations, Nuclear Engineering and Design, 17, 1 (1971) 1. [37] D.A. Canonico, R.G. Berggren, Tensile and Impact Properties of Thick-Section Plate and Weldments, Nuclear Engineering and Design, 17, 1 (1971). [38] F.J. Loss, Effect of Mechanical Constraint on the Fracture Characteristics of Thick-Section Steel, Nuclear Engineering and Design, 17, 1 (1971) 16. [39] P.B. Crosley, E.J. Ripling, Crack Arrest Toughness of Pressure Vessel Steels, Nuclear Engineering and Design, 17, 1 (1971) 32. [40] P.N. Randall, J.G. Merkle, Gross Strain Crack Tolerance of Steels, Nuclear Engineering and Design, 17, 1 (1971) 46. [41] N. Levy, P.V. Marcal, J.R. Rice, Progress in Three-Dimensional Elastic-Plastic Analysis for Fracture Mechanics, Nuclear Engineering and Design, 17, 1 (1971) 64. [42] T.R. Mager, Fracture Toughness Properties of Heavy Section A 533, Grade B, Class 1 Steel Plate and Submergod-Arc Weldment, Nuclear Engineering and Design, 17, 1 (1971) 76. [43] F.J. Witt, T.R. Mager, Fracture Toughness Klcd Values at Temperatures up to 550 ° F for ASTM A 533 Grade B Class 1 Steel, Nuclear Engineering and Design, 17, 1 (1971) 91. [44] R.G. Berggren, W.J. Stelzman, Radiation Strengthening and Embrittlement in Heavy Section Plate and Welds, Nuclear Engineering and Design, 17, 1 (1971) 103. [45] J.R. Hawthorne, Postirradiation Dynamic Tear and Charpy-V Performance of 12-in.-Thick A 533-B Steel Plates and Weld Metal, Nuclear Engineering and Design, 17, 1 (1971) 116. [46] C.W. Hunter, J.A. Williams, Fracture and Tensile Behavior of Neutron-Irradiated A 533-B Pressure Vessel Steel, Nuclear Engineering and Design, 17, 1 (1971) 131. [47 ] D.A. Canonico, Transition Temperature Considerations for Thick-Wall Steel Nuclear Pressure Vessels, Nuclear Engineering and Design, 17, 1 (1971) 149. [48] S.C. Grigory, Testing the Six-Inch Thick Flawed Tensile Specimen for the Heavy Section Steel Technology Program, Nuclear Engineering and Design, 17, 1 (1971) 161.

[49] F.J. Loss and W.S. Pellini, Coupling of Fracture Mechanics and Transition Temperature Approaches to Fracture-Safe Design, NRL Report No. 6913, U.S. Naval Research Laboratory, April 14, 1969. [50] W.S. Pellini and F.J. Loss, Integration of Metallurgical and Fracture Mechanics Concepts of Transition Temperature Factors Relating to Fracture-Safe Design of Structural Steels, NRL Report No. 6900, U.S. Naval Research Laboratory, April 27, 1969. [51] F.J. Witt and R.G. Berggren, Size Effect and Disposition in Impact-Specimen Testing of ASTM A 533 Grade B Steel, Experimental Mechanics (1971) 1. [52] F.J. Witt and T.R. Mager, A Procedure for Determining Bounding Values on Fracture Toughness Kic at any Temperature, presented at the Fifth National Symposium on Fracture Mechanics, University of Illinois, August 31 - September 2, 1971. [53] T.R. Mager, Fatigue Crack Growth Characteristics of Nuclear Pressure Vessel Grade Materials, Westinghouse Electric Corporation, PWR Systems Division (in preparation). [54] F.J. Witt, The Equivalent Energy Method for Calculating Elastic-Plastic Fracture, presented at the Fourth National Symposium on Fracture Mechanics, Pittsburgh, Pennsylvania, August 24-26, ~970 (also USAEC Report ORNL-TM 3172, Oak Ridge National Laboratory). [55] F.J. Witt, The Prediction of the Fracture Behavior of the Large Tensile Tests Using the Equivalent Energy Method, presented at the Fifth Annual Information Meeting of the Heavy Section Steel Technology Program, March 25-26, 1971, Oak Ridge National Laboratory (to be published). [56] F.J. Witt, Fracture Behavior of Pressure Vessel Steel in the Frangible, Transitional and Tough Regimes, presented at the First Internat. Conf. Struct. Mechan. in Reactor Technol., Sept. 20-24 (1971) Berlin, Germany; Nuclear Engineering and Design 20 (1972) 237. [57] R.W. Derby et al., Crack Sharpening by local Fatigue, presented at the Fifth Annual Information Meeting of the Heavy Section Steel Technology Program, March 25-26, 1971, Oak Ridge National Laboratory. [58] D.A. Canonico and J.D. Hudson, Technique for Generating Sharp Cracks in Low-Alloy High-Strength Steels, presented at the Fifth Annual Information Meeting of the Heavy Section Steel Technology Program, March 25-26, 1971, Oak Ridge National Laboratory. [59] R.W. Derby, Preliminary Results of Steel Model Tests, presented at the Fifth Annual Information Meeting of the Heavy Section Steel Technology Program, March 25-26, 1971, Oak Ridge National Laboratory. [60] R.W. Derby, Fracture Studies of Model Pressure Vessels Made from Nuclear Grade Steel, presented at the First Internat. Conf. Struct. Mechan. in Reactor Technol., Sept. 20-24 (1971) Berlin, Germany.