Heavy-Section Steel Technology Research Articles

The role of crack arrest in the evaluation of PWR pressure vessel integrity during PTS transients

The PWR pressurized thermal-shock (PTS) issue, which is concerned with the integrity of the reactor pressure vessel during postulated overcooling transients, is under intensive investigation by the USNRC. The USNRC-sponsored Integrated Pressurized Thermal-Shock (IPTS) and the Heavy-Section Steel Technology (HSST) Programs are dedicated to a better understanding and a timely resolution of the problem. The HSST program is investigating flaw behavior in large cylinders and is also obtaining fracture-mechanics-related material properties, while the IPTS program is primarily concerned with an estimation of the overall frequency of vessel failure and identification of dominant transients and design and operating features contributing thereto for specific nuclear plants. One important component of the IPTS study is a probabilistic fracture-mechanics analysis of the reactor vessel, and a point of particular interest therein is the role of crack arrest in mitigating the consequences of the postulated PTS transients. The HSST program has provided crack-arrest data from small specimens and large thermal-and pressure-loaded cylinders that tend to establish the validity of the crack-arrest concept for application to the PTS problem. Unfortunately, recent results of the IPTS studies indicate that the inclusion of crack arrest in the probabilistic fracture-mechanics model does not substantially influence the calculated frequency of vessel failure. However, there are still significant questions regarding flaw behavior at upper-shelf temperatures, and the HSST program is continuing to pursue this area of uncertainty.

Application of the Failure Assessment Diagram to the Evaluation of Pressure-Temperature Limits for a Pressurized Water Reactor

The failure assessment diagram approach, an elastic-plastic fracture mechanics procedure based on the J-integral concept, was used in the evaluation of pressure-temperature (P-T) limits for the beltline region of the vessel of a pressurized water reactor. The main objective of this paper is to illustrate the application of an alternate fracture mechanics method for the evaluation of pressure-temperature limits, as allowed by the Code of the Federal Regulation 10 CFR 50, Appendix G. The evaluation of P-T limits for the beltline region of a pressurized water reactor was based on the following assumptions: • ASME Pressure Vessel and Piping Code, Section III, Appendix G reference flaw • End-of-life fluence level in the beltline region • Longitudinal flaw in the beltline weld • J-resistance material toughness curves obtained from the U.S. Nuclear Regulatory Commission’s heavy section steel technology (HSST) program • Other material properties obtained from the Babcock & Wilcox Integrated Reactor Vessel Material Surveillance Program The maximum allowable pressure levels were calculated at 33 time points along the given reactor bulk coolant temperature history representing the normal operation of a pressurized water reactor. The results of the calculations showed that adequate margins of safety on operating pressure for the critical weld in the beltline of the pressurized water reactor vessel are assured.

Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed at the Oak Ridge National Laboratory (ORNL) for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along with applications to pressure vessel experiments.

An LEFM Analysis for the Effects of Weld-Repair-Induced Residual Stresses on the Fracture of the HSST ITV-8 Vessel

A fracture mechanics analysis for an intermediate size pressure vessel is described. The vessel was tested as part of the Heavy Section Steel Technology program at Oak Ridge National Laboratory. The purpose of the test was to evaluate the effects of residual stresses on the fracture of the vessel. Residual stresses were due to a half-bead weld repair located near the fabrication seam weld of the vessel. An external part-through flaw was placed in the seam weld near the weld repair. The test was conducted at −23.3°C to cause nonductile fracture behavior and thus accent the effects of residual stresses. The vessel was loaded with internal pressure. Test results showed that crack growth initiated at a significantly lower internal pressure than would be expected if there were no residual stresses. The crack arrested before growing through the thickness of the vessel. The pressure was increased until crack growth was reinitiated. The crack arrested after causing the vessel to leak but not break. The fracture analysis of the test presented here is based on static linear elastic fracture mechanics. Stress intensity factors were evaluated for a range of crack depths using a finite element representation and an energy release rate computation. The effects of residual stresses due to the weld repair were included in the fracture analysis based on residual stress data. There are two important findings of this study. First, good agreement was obtained between the results of the computational fracture analysis and the test data for three characteristics of the test; the pressure at which crack growth initiated, the amount of crack growth at crack arrest and the increase in pressure required to reinitiate crack growth. Thus the observed effects of residual stresses are predictable. The second finding is that the combination of laboratory data and computational results indicate a significant effect of residual stresses on both the crack growth initiation pressure and the crack arrest behavior for these particular test conditions. Estimates are that residual stresses reduced the critical pressure for flaw initiation to 25 percent of the value for the case with no residual stresses. These results are discussed in terms of available data and the computational model.

An Analysis Procedure for Predicting Weld Repair Residual Stresses in Thick-Walled Vessels

An analysis procedure for computing residual stresses due to weld repairs of heavy steel sections is described and illustrated. The model has several unique features for weld repair applications. These include a simple formula for computing temperatures in and near the repair region, a method for economically modeling the application of hundreds of weld passes typically required in repairs, and a method by which the basically two-dimensional residual stress analysis model can be inexpensively modified to incorporate some three-dimensional effects. Laboratory data for residual stresses due to the weld repair of a flat plate and computed values for a curved plate are compared. Good qualitative agreement is found. Next, the computational model is used to predict residual stresses due to a weld repair in one of the Heavy Section Steel Technology Intermediate Test Vessels. Residual stress data available for this vessel are compared with predicted values of residual stress. There is good agreement between the computed values and residual stress data for regions outside of the weld repair. A discussion of the overall agreement is given. This study demonstrates that it is economically feasible to predict the complex behavior of weld repair residual stresses.

Advances in fracture mechanics analyses of primary system performance under operating and accident conditions

Safety research sponsored by the Nuclear Regulatory Commission, Division of Reactor Safety Research, has resulted in notable advances in several areas of importance in the safety evaluation of reactor primary systems under normal operations and accident situations. First, the methods of linear elastic fracture mechanics and of elastic plastic fracture mechanics have been validated for prediction of pressure vessel performance by the Intermediate Vessel Test program results at the Oak Ridge National Laboratory. The test results from both hydraulically and pneumatically loaded pressure vessels are in good agreement with the pretest predictions of vessel failure for the given conditions of pressure, temperature, material properties, and type and size of flaw. This has been verified in base material, both plate and forging, in submerged arc, stress relieved welds, and in weld repair regions. This ability confidently to predict vessel performance under realistic service conditions has permitted development of the computer program OCTAVIA which computes failure curves for a range of flaw sizes in terms of pressure and temperature for specified pressure vessel material at specific neutron fluence levels. It then considers the probability of occurrence of flaw sizes and magnitude of pressure during an operational, overpressurization transient and determines the probability of failure, for both individual flaw sizes and for the full spectrum. A second advance in fracture mechanics technology is related to thermal shock analysis. This advance has been verified by the confirmatory results of testing small thick-walled cylinders under thermal shock conditions in the Heavy Section Steel Technology program, and of warm prestressing tests at the US Naval Research Laboratory. The analyses developed in these efforts have shown that operational pressure-temperature transients can be effectively modeled and applied to pressure vessel materials having appropriate mechanical properties simulating radiation embrittlement, and that the crack initiation/arrest conditions predicted will occur. Thus, the methodology has been developed for the analysis, and it shows continuing integrity of reactor vessels even in the presence of severe thermal shock transients. Thirdly, the technology of crack arrest has reached a level wherein standardization of test specimens and testing methods is now possible and, indeed, is underway. The detailed factors which affect arrest of running cracks, and how those factors must be evaluated to permit determination of a minimum toughness descriptive of crack arrest are better understood. Although this area of fracture mechanics technology still requires much validation testing, the currently available results are already being utilized in operating and accident analyses.

Heavy-section steel technology program: Intermediate-scale pressure vessel tests

Eight flawed 990-mm-OD, 152-mm-thick vessels of ASTM A508 class 2 forging steel or A533, grade B, class 1 steel plate have been pressurized hydraulically to burst or rupture. The rupture test of one vessel (V-7) was repeated pneumatically to study the effects of the sustained load thus attainable. Test temperatures ranged from 0 to 91°C so that transitional to upper shelf fracture toughness behavior was observed. Pretest predictions of failure pressure were based on tensile and toughness properties determined from Charpy-size to 102-mm-thick specimens representative of the test vessels. Linear elastic fracture mechanics based on strain and plastic instability analyses were generally adequate for determining the failure pressure, which ranged from 2.15 to 3.28 times design pressure.

The USAEC heavy section steel technology program: Objectives and status

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.

Postirradiation dynamic tear and Charpy-V performance of 12-in. thick A533-B steel plates and weld metal

Charpy-V (C V) and dynamic tear (DT) test comparisons are developed for the irradiated condition of two 12 in. thick A533-B steel plates and a submerged arc weld deposit. Materials for the investigation were donated by the U.S. Atomic Energy Commission's Heavy Section Steel Technology (HSST) Program. Individual plates were identified as HSST plates 01 and 02; the weld was identified as weld 50. Low temperature (<300°F, 149°C) and elevated temperature (550°F, 288°C) irradiations were conducted. Fluences for the eight irradiation experiments ranged from 2.5–3.3 × 10 19 n/cm 2 (> 1 MeV). Postirradiation C V and DT mid-energy transition temperature increases were found comparable for both plate and weld deposit (quarter thickness position evaluations). For the case of plate, percentage reductions in C V and DT shelf energy with irradiation also compared well. As a result, preirradiation DT to C V shelf energy ratios appeared to be retained with irradiation. Shelf observations are in agreement with recent findings for A543, A302-B, and other A533 steel plates which illustrate a DT to C V shelf energy ratio of 8.0:1 for pre- and postirradiation conditions. It is concluded that postirradiation DT properties of A533-B plate, including mid-energy transition and shelf level, can be calculated with reasonable accuracy from postirradiation C V results for use in fracture safety analyses. Assessments of shelf level toughness before and after irradiation are developed by ratio analysis diagram (RAD) procedures. Unlike observations for plate, a significant reduction in DT to C V shelf energy ratio by irradiation was found for the weld deposit. Weld metal fracture appearances suggest preferred paths for fracture within the duplex (layered) microstructure. Radiation embrittlement sensitivity at 550°F (288°C) was found to increase with increasing copper content as expected; however, embrittlement sensitivities of all three materials exceeded expectations. An influence of nickel content on radiation embrittlement sensitivity development is proposed.

Testing the six-inch thick flawed tensile specimen for the heavy section steel technology program

The 15 million pound tensile test facility at Southwest Research Institute has been used to test a series of large tensile specimens for the Heavy Section Steel Technology (HSST) Program. These specimens are A533 grade B, class 1, and are 6 in thick, 18 in wide, 4 ft long and contain flaws. The procedures used to create the flaws are discussed and pertinent data for six tests at temperatures from 50°F to 220°F and requiring loads up to 7.5 million pounds for failure are presented.

Abstracts of papers presented at the fifth annual information meeting of the heavy section steel technology program : Oak Ridge National Laboratory, March 25–26, 1971

Abstracts of papers presented at the fifth annual information meeting of the heavy section steel technology program : Oak Ridge National Laboratory, March 25–26, 1971

Fracture toughness properties of heavy section A533, grade B, class 1 steel plate and submerged arc weldment

This paper presents a review and summary of the fracture toughness ( K I c ) properties generated for A533 grade B class 1 steel plate and submerged arc weldment material. The materials for which the data are presented were specially evaluated in the Heavy Section Steel Technology Program. The fracture toughness of HSST plate and weldment material is presented as a function of loading rate, metallurgical condition and irradiation (fluence). The data indicate the fracture toughness properties to be highly temperature dependent.

Abstracts of papers presented at the Fourth Semiannual information meeting of the Heavy Section Steel Technology Program : Oak Ridge National Laboratory, March 31 and April 1, 1970

Abstracts of papers presented at the Fourth Semiannual information meeting of the Heavy Section Steel Technology Program : Oak Ridge National Laboratory, March 31 and April 1, 1970

The heavy section steel technology program an engineering investigation of the structural behavior of thick pressure vessels

The Heavy Section Steel Technology Program is a United States Atomic Energy Commission sponored effort which was initiated to study the structural behavior of the thick section plates and pressure vessels fabricated from such plates for use in light water power reactor systems. The major emphasis is on the effects of flaws and material inhomogeneities and discontinuities under in-service reactor operating and accident conditions. The fracture modes of these vessels are of special interest because of the dangers peculiar to nuclear reactors should a major failure of a pressure vessel occur while in service. The program is being carried out in very close cooperation with the materials, fabrication, and design interests of the United States nuclear power industry. The scope of the program is outlined and the objectives of the various tasks are identified. Some results obtained during the initial efforts of the program are presented.

3 Solar water heating in Great Britain

As part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel. This boiling water reactor vessel was never in service. One objective was to evaluate the approximate 0.7- by 3-m (2- by 10-ft) segment of the vessel provided using ultrasonic flaw detection methods performed with both ASME Code techniques and supplemental ultrasonic methods. A second objective was to evaluate the inner surface stainless steel cladding for cracks with a high sensitivity penetrant examination. Both objectives were successfully completed. Five Code-recordable indications were detected ultrasonically; however, all were found to be anomalies associated with the cladding. One flaw was detected by the supplemental ultrasonic tests, and it was analyzed destructively. This flaw was pipelike indication, about 20 mm (0.8 in.) long extending along the length of the longitudinal weld in which it was located and was about 20 mm below the cladding surface. The flaw had a through-wall dimension (or length) of about 6 mm (0.24 in.) for an approximate 3-mm (0.1-in.) distance along the 20-mm major length. No flaws were detected by the penetrant examination of the cladding surface.