Heavy-Section Steel Technology Program Research Articles

Heavy-Section Steel Technology and Irradiation Programs—Retrospective and Prospective Views

In 1965, the Atomic Energy Commission, at the advice of the Advisory Committee on Reactor Safeguards, initiated the process that resulted in the establishment of the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory. In 1989, the Heavy-Section Steel Irradiation (HSSI) program, formerly the HSST task on irradiation effects, was formed as a separate program, and in 2007, the HSST/HSSI programs, sponsored by the U.S. Nuclear Regulatory Commission (USNRC), celebrated 40 years of continuous research oriented toward the safety of light-water nuclear reactor pressure vessels (RPVs). This paper presents a summary of results from those programs with a view to future activities. The HSST program was established in 1967 and initially included extensive investigations of heavy-section low-alloy steel plates, forgings, and welds, including metallurgical studies, mechanical properties, fracture toughness (quasi-static and dynamic), fatigue crack-growth, and crack-arrest toughness. Also included were irradiation effects studies, thermal shock analyses, testing of thick-section tensile and fracture specimens, and nondestructive testing. In the subsequent decades, the HSST Program conducted extensive large-scale experiments with intermediate-size vessels (with varying size flaws) pressurized to failure, similar experiments under conditions of thermal shock and even pressurized thermal shock (PTS), wide-plate crack-arrest tests, and biaxial tests with cruciform-shaped specimens. Extensive analytical and numerical studies accompanied these experiments, including the development of computer codes such as the recent Fracture Analysis of Vessels Oak Ridge code currently being used for PTS evaluations. In the absence of radiation damage to the RPVs, fracture of the vessel is improbable. However, it was recognized that exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The HSSI Program conducted a series of experiments to assess the effects of neutron irradiation on RPV material behavior, especially fracture toughness. These studies included RPV plates and welds, varying chemical compositions, and fracture toughness specimens up to 101.6 mm (4 in.) thickness. 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. Results from the HSST and HSSI programs are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety.

Weibull statistical models of KIc/ KIa fracture toughness databases for pressure vessel steels with an application to pressurized thermal shock assessments of nuclear reactor pressure vessels

This paper presents new statistical representations of recently extended fracture toughness K Ic and K Ia databases for pressure vessel steels. These models were developed by the Heavy Section Steel Technology program at Oak Ridge National Laboratory in support of the current effort by the U.S. Nuclear Regulatory Commission to update its regulatory guidance for pressurized-thermal-shock (PTS) transients in nuclear reactor pressure vessels. The Weibull distribution, with two of its parameters calculated by the Method of Moments point-estimation technique, forms the basis for the new statistical models. An application of the new K Ic/ K Ia models, as implemented in the favor probabilistic fracture mechanics computer program, is also presented for three PTS transients.

Shallow Flaws Under Biaxial Loading Conditions—Part II: Application of a Weibull Stress Analysis of the Cruciform Bend Specimen Using a Hydrostatic Stress Criterion1

Cruciform beam fracture mechanics specimens have been developed in the Heavy Section Steel Technology Program at Oak Ridge National Laboratory to introduce a prototypic, far-field, out-of-plane biaxial bending stress component in the test section that approximates the nonlinear biaxial stresses resulting from pressurized-thermal-shock or pressure-temperature loading of a nuclear reactor pressure vessel (RPV). Matrices of cruciform beam tests were developed to investigate and quantify the effects of temperature, biaxial loading, and specimen size on fracture initiation toughness of two-dimensional (constant-depth) shallow surface flaws. Tests were conducted under biaxial load ratios ranging from uniaxial to equibiaxial. These tests demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. Two and three-parameter Weibull models have been calibrated using a new scheme (developed at the University of Illinois) that maps toughness data from test specimens with distinctly different levels of crack-tip constraint to a small-scale-yielding Weibull stress space. These models, with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, have been shown to correlate the experimentally observed biaxial effect in cruciform specimens, thereby providing a scaling mechanism between uniaxial and biaxial loading states.

Heavy-section steel technology program: Recent developments in crack initiation and arrest research

Technology for the analysis of crack initiation and arrest is central to the reactor pressure vessel fracture-margin-assessment process. Regulatory procedures for nuclear plants utilize this technology to assure the retention of adequate fracture-prevention margins throughout the plant operating license period. As nuclear plants age and regulatory procedures dictate that fracture- margin assessments be performed, interest in the fracture-mechanics technology incorporated into those procedures has heightened. This has led to proposals from a number of sources for development and refinement of the underlying crack-initiation and arrest-analysis technology. This paper presents an overview of ongoing Heavy-Section Steel Technology (HSST) Program research aimed at refining the fracture toughness data used in the analysis of fracture margins under pressurized-thermal-shock loading conditions.

Heavy-section steel technology program overview

This paper presents a status review of ongoing HSST program tasks aimed at refining the technology used in analysis of reactor pressure vessel fracture margins under pressurized thermal-shock (PTS) loading. Specific fracture-technology issues addressed include vessel flaw density and distribution, shallow flaws, fractures-toughness data transfer, circumferential cracks, ductile tearing and the influence of low-tearing toughness in stainless steel cladding. Preliminary results from the analysis and test programs are presented, together with interim assessments of their potential impact on a reactor vessel PTS analysis.

Heavy-Section Steel Technology Program: Fracture Issues (Survey Paper)

Large-scale fracture mechanics tests have resulted in the identification of a number of fracture technology issues. Identification of additional issues has come from the reactor vessel materials irradiation test program and from reactor operating experience. This paper provides a review of fracture issues with an emphasis on their potential impact on a reactor vessel pressurized thermal shock (PTS) analysis. Mixed mode crack propagation emerges as a major issue, due in large measure to the poor performance of existing models for the prediction of ductile tearing. Rectification of ductile tearing technology deficiencies may require extending the technology to include a more complete treatment of stress state and loading history effects. The effect of cladding on vessel fracture remains uncertain to the point that it is not possible to determine at this time if the net effect will be positive or negative. Enhanced fracture toughness for shallow flaws has been demonstrated for low-strength structural steels. Demonstration of a similar effect in reactor pressure vessel steels could have a significant beneficial effect on the probabilistic analysis of reactor vessel fracture. Further development of existing fracture mechanics models and concepts is required to meet the special requirements for fracture evaluation of circumferential flaws in the welds of ring-forged vessels. Fracture technology advances required to address the issues discussed in this paper are the major objective for the ongoing Heavy Section Steel Technology (HSST) program at ORNL.

Heavy-section steel irradiation program summary

In light-water-cooled nuclear power plants, a steel reactor pressure vessel (RPV) forms the primary containment boundary for the nuclear fuel assembly. Since a failure of the RPV carries the potential of major contamination release and severe accident, it is imperative to safe reactor operation to understand and be able to accurately predict realistic failure models of the vessel material. A major known risk arises from irradiation-induced degradation of the RPV fracture resistance which occurs as a result of normally occurring neutron flux during the service life of the reactor. Better understanding of the irradiation degradation of the vessel material will provide better input to improve the engineering models vital for prediction of RPV integrity and safe service life. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water RPVs. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness ( K 1c and J 1c), crack-arrest toughness ( K 1a), ductile tearing resistance (d J/da), Charpy V-notch impact energy, dropweight nil-ductility temperature (NDT), and tensile properties. Models based on observations of radiation-induced microstructural changes using field ion and high-resolution transmission electron microscopy provide a firmer basis for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI Program are highcopper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. In addition, a limited effort Will focus on stainless steel weld overlay cladding, typical of that used on the inner surface of RPVs, since its postirradiation fracture properties have the potential for strongly affecting the extension of small surface flaws during overcooling transients. Results from the HSSI studies will be integrated to aid in resolving major regulatory issues facing the U.S. Nuclear Regulatory Commission which involve RPV irradiation embrittlement such as pressurized-thermal shock, operating pressure-temperature limits, low-temperature overpressurization, and the specialized problems associated with low upper-shelf welds. Taken together the results of these studies also provide guidance and bases for evaluating both the aging behavior and the potential for plant life extension of light-water RPVs.

Viscoplastic modeling of dynamic fracture events in reactor pressure vessel steels

The Heavy-Section Steel Technology Program at the Oak Ridge National Laboratory under the sponsorship of the U.S. Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel steels. Nonlinear rate-dependent effects in crack run-arrest events in ductile materials are being investigated through the development and application of viscopalstic-dynamic finite-element analysis techniques. This paper describes a portion of these studies wherein various viscoplastic constitutive models, dynamic crack-propagation algorithms, and proposed nonlinear fracture criteria are being evaluated through their use in the general-purpose ADINA finite-element computer program. The predictive capabilities of the nonlinear techniques are being examined by comparing analytical and experimental results for the crack-arrest tests of large nonisothermal plate specimens. Values of the fracture parameters calculated by elastodynamic and viscoplastic dynamic techniques are compared to assess the impact of including visocplastic effects in the computational models. Recent studies are described that examine the convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for the fracture parameters in the context of viscoplastic-dynamic models are discussed.

Results of irradiated cladding tests and clad plate experiments

Two aspects critical to the fracture behavior of three-wire stainless steel cladding were investigated by the Heavy-Section Steel Technology (HSST) Program: (1) radiation effects on cladding strength and toughness; and (2) the response of mechanically loaded, flawed structures in the presence of cladding (clad plate experiments). Postirradiation testing results show that, in the test temperature range from −125 to 288°C, the yield strength increased, and ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing. Radiation damage decreased the Charpy upper-shelf energy by 15 to 20% and resulted in up to 28°C shifts of the Charpy impact transition temperature. Results of irradiated 12.5 mm-thick compact specimens (0.5TCS) show consistent decreases in the ductile fracture toughness, J Ic , and the tearing modulus. Results from clad plate tests have shown that: (1) a tough surface layer composed of cladding and/or heat-affected zone has arrested running flaws under conditions where unclad plates have ruptured; and (2) the residual load-bearing capacity of clad plates with large subclad flaws significantly exceeded that of an unclad plate.

Performance of low-upper-shelf material under pressurized-thermal-shock loading

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. The experiment was performed in the pressurized-thermal-shock test facility at the Oak Ridge National Laboratory. PTSE-2 was designed primarily to reveal the interaction of ductile and brittle modes of fracture and secondarily to investigates the effects of warm prestressing, A test vessel was prepared by inserting a cracklike flaw of well-defined geometry on the outside surface of the vessel. The flaw was 1 m long by ≈ 15 mm deep. The instrumented vessel was placed in the 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 twoseparate transients, each one starting with the vessel nearly isothermal. The first transient induced a warm-prestressed state, during which K I , first exceeded K 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.

Flaw density examinations of a clad boiling water reactor pressure vessel segment

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.

Late-event viscoplasticity in wide-plate crack-arrest tests

A primary objective of the crack-arrest studies being conducted by the Heavy-Section Steel Technology (HSST) program is to understand pressure vessel conditions that would initiate growth of an existing crack and conditions that would lead to arrest of a moving crack. In meeting this objective, the HSST program is generating crack-arrest data over an expanded temperature range through tests involving large cylinders, pressure vessels and wide-plate specimens. The role of nonlinear rate-dependent effects in the interpretation of crack run-arrest events in ductile material is being investigated by the HSST program through development and applications of viscoplastic-dynamic finite element analysis techniques. This paper describes the portion of those studies that relate to the installation of two viscoplastic constitutive models and several proposed fracture criteria into the ADINA finite element program. The formulations of the Bodner-Partom and the Perzyna constitutive models that were installed in ADINA are presented, including a compilation of material parameters for these models obtained from dynamic stress-strain data for A533 grade B class 1 steel. This is followed by a description of path-area integrals J′, T ∗ (from Nishioka and Atluri) and J ̂ (from Kishimoto) and of the function γ (representing energy flow to the crack tip) installed in the ADINA program. Finally, the combined predictive capabilities of these techniques are applied to the analysis of one of the recently completed HSST wide-plate crack-arrest tests.

Wide-plate crack-arrest tests utilizing a prototypical pressure vessel steel

Wide-plate crack-arrest tests are being performed at the National Bureau of Standards (Gaithersburg, MD) under the Heavy-Section Steel Technology (HSST) Program and are designed to provide fracture-toughness measurements at temperatures approaching or above the onset of the upper-shelf regime, in a rising toughness region and with increasing driving force. The test specimens are 1 × 1 × 0·1 m and possess a single-edge notch (crack) that initiates in cleavage propagation at low temperature and arrests in a region of increased fracture toughness. The toughness is achieved through a linear transverse temperature profile across the plate. Results obtained using a prototypical reactor pressure vessel steel (A533 grade B class 1 material) exhibit a significant increase in toughness at temperatures near and above the onset of Charpy upper shelf. Additionally, cleavage crack propagation and arrest at temperatures above the onset of Charpy upper shelf have been demonstrated.

HSST crack-arrest studies overview

An overview is given of the efforts underway in the Heavy-Section Steel Technology (HSST) Program as of October 1985 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 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 analysis 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.

Some advances in fracture studies under the heavy-section steel technology program

Recent results are summarized from HSST studies in three major areas that relate to assessing nuclear reactor pressure vessel integrity under pressurized-thermal-shock (PTS) conditions. These areas are irradiation effects on the fracture properties of stainless steel cladding, crack run-arrest behavior under non-isothermal conditions, and fracture behavior of a thick-wall vessel under combined thermal and pressure loadings. Since a layer of tough stainless steel weld overlay cladding on the interior of a pressure vessel could assist in limiting surface crack extension under PTS conditions, its resistance to radiation embrittlement was examined. A stainless steel overlay cladding, applied by a submerged arc, single-wire, oscillating-electrode method, was irradiated to 2 × 10 23 neutrons/m 2 (> 1 MeV) at 288°C. Yield strength increases up to 27% and a slight increase in ductility were observed. Charpy V-Notch data showed a ductile-to-brittle transition behavior caused by temperature-dependent failure of the 8-ferrite phase. The type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation. Crack-arrest behavior of A533 grade B class 1 steel was examined for temperatures extending above the onset of Charpy upper-shelf. Crack-arrest experiments that use wide-plate specimens have shown crack arrest occurring prior to transition to tearing or tensile instability. High values of crack-arrest toughness have been recorded (static values above 400 MPa m that are well above the maximum value that safety assessment criteria assume such materials can exhibit. A validation experiment was performed by exposing an intentionally flawed HSST intermediate test vessel to combined pressure and thermal transients. The experiment addressed warm-prestressing phenomena, crack propagation from brittle to ductile regions, and crack stabilization in ductile regions. Test and analysis results are summarized.

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.

The heavy-section steel technology pressurized-thermal-shock experiment, PTSE-1

A pressurized-thermal-shock (PTS) facility was developed in the Heavy-Section Steel Technology Program at Oak Ridge National Laboratory for performing experiments that challenge predictions of analytical methods applicable to full-scale reactor pressure vessels under combined loading. The first experiment (PTSE-1) was designed to address three principal issues: (1) warm-prestressing phenomena; (2) crack propagation from brittle to ductile regions; and (3) transient crack stabilization in ductile regions. The paper presents a description of the PTS facility at ORNL and a review of the objectives and results of the first test. Also included are elastodynamic finite-element analyses of the two crack run-arrest events that occurred in the second and third phases of the test. Finally, some conclusions and recommendations are presented based on the outcome of the first experiment.

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.

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.

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.