Hot Isostatic Pressing (HIP) of type 316L austenitic stainless steel powder has been an established manufacturing practice for more than twenty five years within the oil and gas sectors and more recently in the naval defence industry. The successful ASME Code Case approval (N-834) has facilitated the manufacture of 316L components via Powder Metallurgy HIP (PM/HIP) for the civil nuclear sector. However, a number of issues have tended to hinder the uptake of PM/HIP as an alternative viable manufacturing route for both castings and forgings. Firstly, the powder specification for 316L and HIP processing parameters has typically been left to the discretion of the manufacturers. As such, the finer details of HIP product specification require greater clarity and definition for optimum performance/reproducibility. Secondly, comparison of historical data for 316L PM/HIP has shown variation in the Charpy impact toughness performance. These differences have been attributed to the oxygen content of the atomised powder, with greater oxygen contents yielding product with reduced impact properties. Based on these factors, a systematic study of the current state of the art of 316L commercial powder production, encapsulation/consolidation and selected HIP parameters was undertaken in collaboration with the Electric Power Research Institute (EPRI). A 316L powder specification was developed that primarily limited the oxygen content of the powder to under 130ppm. This lower oxygen limit reflects the improvements that commercial powder suppliers have been making over the past decade to ensure greater powder cleanliness. The test programme generated a significant body of test data based on 3 × 3 × 3 matrix of: powder supply, HIP service provider and HIP sustain times. The results were excellent across the full range of variables studied with all test billets passing the specification requirements of ASTM A988 and additional imposed requirements. Very consistent 316L material properties were produced for billets manufactured via differing HIP service providers across the comprehensive destructive test programme. This demonstrates the robustness and uniformity of the PM/HIP supply chain in producing 316L material of the requisite quality. In addition, no significant difference in material properties was noted for material pressed between 2–8 hours hold time, suggesting that the HIP process window is large with respect to hold time. Of significant note was that material produced with one powder yielded material with consistently the highest strengths and Charpy impact toughness. This has been attributed to chemical composition of the powder, which featured both a low oxygen and also a high nitrogen content.

The selection of cutting tool geometry and cutting conditions plays a major part in the planning of machining operations affecting the productivity. Although there are several classifications used by tool manufacturers, carbide inserts for face milling can be categorised according to their cutting edge angle. In this study, the material removal performance of square, round and high-feed inserts, which all have different cutting edge angles, was investigated through cutting trials conducted without coolant on forged SA508 Gr.3 Cl.2. Predictive tool life models for round and high-feed inserts were developed by means of the response surface methodology and their adequacy was checked through the analysis of variance. Cutting speed versus feed per tooth contour plots are presented against predicted tool life values which facilitate the selection of cutting conditions ensuring desired tool life. The experimental data has demonstrated that using inserts with lower cutting edge angle would significantly improve cycle times especially in the rough milling stage where the surface integrity is not of primary interest. However, it should be taken into account that limitations exist mainly due to the process stability and capabilities of machine tools.

Following the flaw indications found in summer 2012 in two Belgian Reactors Pressure Vessels (RPV), WENRA recommended [1] the nuclear safety authorities in Europe to verify the material quality and integrity of the RPV in a 2-step approach: 1) a comprehensive review of the manufacturing and inspection records of the forgings of the RPV, 2) an additional UT examination of the base material of the vessels if needed. In this context, and to consolidate scientific basis on this issue, IRSN, the French technical safety organization, conducted, with CEA support, a test program aiming at studying the consequences of hydrogen flakes in large forgings of primary equipment (RPV, steam generator, pressurizer). Framatome provided the material to be investigated, namely two blocks of a steam generator vessel shell in 18MND5 steel: a block without flake — the reference block — and a block including a high density of hydrogen flakes. This shell — so called VB395 — was rejected because of an incident which occurred during the degassing heat treatment. Fracture toughness has been evaluated from 85 tests in the ductile range and the ductile-to-brittle transition range of the material. The test results on usual 0.5T-CT specimens were compared to those on specimens containing a hydrogen flake replacing the fatigue precrack. The latter were interpreted using 3D elastic-plastic X-FEM simulations allowing the modelling of the irregular flake geometry. Furthermore, large scale bending specimens with multiple flakes have been tested at −100°C. These tests were interpreted thanks to 3D X-FEM simulations allowing the analysis of the hydrogen flake interaction in terms of K J .

Regions of higher-than-normal carbon content due to carbon macro-segregation have been found in large, pressure retaining forged ferritic steel components in some nuclear reactors. Higher carbon content in ferritic steel can decrease the resistance to fracture from the presence of flaws in the material. Acceptable margins against failure of pressurized components in nuclear safety systems must be maintained throughout their service life to ensure core integrity for all operational and postulated transient loading events. Should carbon macro-segregation substantially reduce the material resistance to fracture in safety components, then the margins against through-wall flaw propagation may fall below those specified by regulatory requirements to ensure adequate component and reactor core integrity. Probabilistic fracture mechanics (PFM) analyses were performed to assess the risk and structural significance of postulated carbon macro-segregation in large, forged pressure retaining components in pressurized water reactors (PWRs). The risk assessment was performed to evaluate several forged components and two classes of loading events. The forged components include the ring and head forgings in the reactor pressure vessel (RPV), steam generator (S/G) and pressurizer. The loading events used in the risk evaluation include pressurized thermal shock (PTS) transient events and a normal RPV cooldown event. The analyses included a range of component dimensions, surface and embedded flaw distributions, various levels of carbon macro-segregation up to and beyond the maximum measured values for the components, and the effects of neutron irradiation, including the effects of potential copper and phosphorus co-segregation. The PFM analyses were performed using the software, Fracture Analysis of Vessels, Oak Ridge (FAVOR). The results from the risk assessment indicate that: acceptable margins against failure are maintained through an 80-year operating interval even if carbon macro-segregation were to be present in RPV, S/G and pressurizer ring and head forgings in PWRs; and the risk associated with the presence of carbon macro-segregation in PWR ring and head forgings is significantly lower than regulatory risk related acceptance criteria.

Recent upstream oil and gas industry experience has raised attention to substandard properties with high strength carbon steel forgings manufactured to the requirements of ASTM A694 and MSS-SP-44. As part of an internal investigation into quality of commodity pipeline flanges, three flanges certified as ASTM A694 grade F60 to F70, were purchased off-the-shelf from three different manufacturers for microstructural and mechanical property investigation. All three flanges were supplied with material test certificates indicating acceptable material properties. Tensile and Charpy impact specimens were extracted from various locations and orientations in each flange. All three flanges failed to meet yield strength requirements for the specified grade. The impact energy and shear area values were well below those reported on the material test certificates. The discrepancy between the sacrificial testing results and the material test certificates is attributed to the use of separately forged test blocks for quality testing instead of integral prolongations or a sacrificial production part, which is permitted by ASTM A694 and MSS-SP-44. Further investigation was made into the chemical composition and heat treating practices. The chemical composition can be characterized as high strength, low alloy steel (HSLA) by virtue of 0.05–0.08 wt. pct. vanadium added to a carbon-manganese steel with CE(IIW) ranging from 0.43 to 0.45. Advanced microscopy showed that the morphology of the vanadium precipitates was inadequate as a strengthener and deleterious to Charpy impact properties for the size of the flanges and the heat treatment practices applied.

The new ferritic heat-resistant steel composed of 9Cr-3W-3Co-Nd-B, registered as ASME Code Case 2839, has been developed for large diameter and heavy wall thickness pipes and forgings of fossil-fired power boilers. The steel, which contains 0.01 mass% boron, a small amount of neodymium, and optimized amounts of nitrogen, is characterized by the superior long-term creep strengths of both the base metal and welded joint. P92 had equiaxed subgrain structures changed from martensite lath structures and coarsened M 23 C 6 type carbides after long-term creep. In contrast, the developed steel, SAVE12AD, maintained martensite lath structures with fine M 23 C 6 along the boundaries even after the long-term creep stage. The addition of high amounts of boron suppressed the coarsening of M 23 C 6 along the boundaries, thereby stabilizing the martensite lath structure in the base metal of the steel. Consequently, SAVE12AD had higher creep rupture strength than other high chromium ferritic steels. To investigate the creep rupture strength of welded joints, two welded joints with Ni-based alloy and Grade 92 welding filler wires were prepared by automatic gas tungsten arc welding. The creep rupture strength of each welded joint showed small degradation compared with the base metal in the long-term creep stage over 10,000 hours. These were ruptured 1.5 mm away from the fusion line, which was the same area as Type IV cracking. Microstructural observations were carried out by electron back scatter diffraction analysis using simulated heat-affected zone samples at different peak temperatures from 750 °C to 1350 °C in order to clarify the microstructure in the heat-affected zone. No fine grain area was observed in the microstructure after the simulated heat-affected zone at 910 °C just above A C3 transformation temperature, although there were fine grains along prior austenite grain boundaries, which seemed to form with the diffusion transformation. The creep cracks seemed to have initiated from the fine grain structures, resulting in the rupture at the same area as Type IV cracking. However, the creep rupture strength degradation of the welded joints against the base metal was significantly smaller than that of conventional steel welded joints owing to the suppression of fine grains found in the heat-affected zone heated around A C3 temperature. The developed 9Cr-3W-3Co-Nd-B steel (SAVE12AD) will be used for large diameter and heavy wall thickness pipes and forgings in 600 °C ultra super critical power plants.

The current approach in evaluating the Pressurized Water Reactor (PWR) inlet and outlet nozzle corner regions with respect to plant heat-up and cool-down pressure-temperature limit curves contains a number of conservatisms. These conservatisms include postulation of a large ¼ thickness flaw at the nozzle corner region and use of RT NDT (reference nil-ductility temperature) or an estimation of RT NDT . The paper herein discusses generic fracture toughness of nozzle forging material SA-508 Class 2 for use with postulated smaller surface flaws in developing pressure-temperature limit curves for nozzle corners for nuclear power plant operations. ASME Appendix G uses the lower bound K IC curve, which has inherent margin since RT NDT is a conservative method for locating the K IC curve. RT NDT is based on the drop weight test, which is a crack arrest transition temperature measurement, and the Charpy impact test, which is a blunt notch impact test. These data are conservatively bounded by the K IC curve, which is a lower bound crack initiation toughness curve. In contrast, the master curve method is based on an initiation transition temperature fracture toughness test technique per ASTM E1921. The master curve index temperature (T 0 ) provides a more accurate measure of the material fracture toughness than K IC indexed with RT NDT . Since many of the nuclear pressure vessels were fabricated to ASME Code editions prior to 1972, RT NDT was not measured for the nozzles. In many cases, RT NDT has been estimated. Therefore, for this work, the fracture toughness was generically established based on conservative T 0 measurements of 22 representative forgings with a margin of two standard deviations to ensure a conservative lower bound toughness using ASME Appendix G, G-2110. The properties of a forging are better near the surface due to the faster cooling rate during heat treatment. The difference in reactor pressure vessel fracture toughness was established for forgings near the surface at the postulated flaw location as allowed by ASME Section III, NB-2223.2 relative to the traditional ¼ thickness location. The near-surface forging toughness was conservatively determined through evaluation of 31 near-surface and approximate ¼ thickness location fracture toughness measurements.

This paper provides an overview of the significant revisions pending for the upcoming 2017 edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section VIII Division 3, Alternative Rules for Construction of High Pressure Vessels , as well as potential changes to future editions under consideration of the Subgroup on High Pressure Vessels (SG-HPV). Changes to the 2017 edition include the removal of material information used in the construction of composite reinforced pressure vessels (CRPV); this information has been consolidated to the newly-developed Appendix 10 of ASME BPVC Section X, Fiber-Reinforced Plastic Pressure Vessels . Similarly, the development of the ASME CA-1, Conformity Assessment Requirements standard necessitated removal of associated conformity assessment information from Section VIII Division 3. Additionally, requirements for the assembly of pressure vessels at a location other than that listed on the Certificate of Authorization have been clarified with the definitions of “field” and “intermediate” sites. Furthermore, certain design related issues have been addressed and incorporated into the current edition, including changes to the fracture mechanics rules, changes to wires stress limits in wire-wound vessels, and clarification on bolting and end closure requirements. Finally, the removal of Appendix B, Suggested Practice Regarding Post-Construction Requalification for High Pressure Vessels, will be discussed, including a short discussion of the new appendix incorporated into the updated edition of ASME PCC-3, Inspection Planning Using Risk Based Methods . Additionally, this paper discusses some areas in Section VIII Division 3 under consideration for improvement. One such area involves consolidation of material models presented in the book into a central area for easier reference. Another is the clarification of local strain limit analysis and the intended number and types of evaluations needed for the non-linear finite element analyses. The requirements for test locations in prolongations on forgings are also being examined as well as other material that can be used in testing for vessel construction. Finally, a discussion is presented on an ongoing debate regarding “occasional loads” and “abnormal loads”, their current evaluation, and proposed changes to design margins regarding these loads.

In the summer of 2012, the detection of a large number of quasi-laminar flaw indications in the reactor vessel beltline ring forgings of two Belgian pressurized water reactors (Doel Unit 3 and Tihange Unit 2) posed a significant safety threat that led the licensee to shutdown both plants. Those indications were identified by the licensee as hydrogen flakes that developed during the fabrication of the forgings. As a prerequisite for a potential restart of the units, the Belgian Nuclear Safety Regulator, the Federal Agency for Nuclear Control (FANC), requested the licensee to provide, for each unit, a safety case demonstrating the acceptability of the reactor pressure vessel for continued operation. As the technical subsidiary of the FANC, Bel V performed a safety evaluation of the condition of the reactor pressure vessels. The paper documents the approach Bel V used in his safety evaluation and the criteria he defined to evaluate the acceptability of the hydrogen flaking damage in the reactor pressure vessels.

During 2012 RPVs inspections, nearly-laminar indications were detected in the lower and upper core shells of Doel 3 and Tihange 2 RPVs. As a consequence, the Doel 3 and Tihange 2 NPPs have to stay in cold shutdown until it was proved that they can be safely operated. The performance and capability of the applied UT technique to detect and characterize hydrogen flakes had to be confirmed. To achieve this objective, extensive investigations were launched on blocks extracted from a steam generator shell known for containing hydrogen flakes. The paper describes the UT validation performed and the results of the destructive tests which enable to demonstrate the capability of a straight ultrasonic beam to detect and size nearly laminar indications and to correctly identify ligaments between two adjacent indications. It addresses the formal extension of qualification performed by correlating UT measurements acquired on a large numbers of real hydrogen flakes, simulations of UT responses and destructive examinations. Lessons learned from the experiments are finally described.

The formation of hydrogen flakes in Reactor Pressure Vessels (RPV’s) forgings is a well known phenomenon that can affect forging components during manufacturing. Following recent data from Belgium, it has been the object of important assessments during the last 3 years. We will present in this paper elements pertaining to the RPV’s of the French nuclear fleet. A thorough review of fabrication processes and specifically of ultrasonic examination of forgings, including the oldest ones has demonstrated that the risk of leaving undetected hydrogen flakes in RPV’s forgings is not a problem in French RPVs. The French Regulator also clearly shared this position point in 1985. Parts with defects were once again observed at the beginning of 2012 but the defects were attributed to errors in hydrogen measurement at the steel maker shop. Those hydrogen induced defects were detected at an early manufacturing stage (before quality heat treatment) by ultrasonic examination, and the concerned parts were consequently rejected through the normal quality control process.

In forged, welded, and machined components, residual stresses can form during the fabrication process. These residual stresses can significantly alter the fatigue and fracture properties compared to an equivalent component containing no residual stress. When performing lifetime assessment, the residual stress state must be incorporated into the analysis to most accurately reflect the initial condition of the component. The focus of this work is to present the computational and experimental tools that we are developing to predict and measure the residual stresses in stainless steel for use in pressure vessels. The contour method was used to measure the residual stress in stainless steel forgings. These results are compared to the residual stresses predicted using coupled thermo-mechanical simulations that track the evolution of microstructure, strength and residual stress during processing.

This paper is to give an overview of the major revisions pending in the upcoming 2015 edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section VIII Division 3, Alternative Rules for Construction of High Pressure Vessels , and potential changes being considered by the Subgroup on High Pressure Vessels (SG-HPV) for future editions. This will include an overview of significant actions which will be included in the upcoming edition. This includes action relative to test locations in large and complex forgings, in response to a report from the U.S. Chemical Safety and Hazard Investigation Board (CSB) report of a failed vessel in Illinois. This will also include discussion of a long term issue recently completed on certification of rupture disk devices. Also included will be a discussion of a slight shift in philosophy which has resulted in the linear-elastic stress analysis section being moved to a Non-Mandatory Appendix and discussion of potential future of linear-elastic stress analysis in high pressure vessel design.

316FR and Mod.9Cr-1Mo steels are candidates as structural materials for Japanese Sodium Fast Reactor (JSFR). The design life and operation temperature of JSFR is 60 years and 550 °C, respectively. Time-dependent allowable stress is essential. The evaluation of allowable stresses to 500,000 h is a considerable item. Long term strength is evaluated from a viewpoint of microstructural evaluation related to fracture mechanism. In addition, degradation after long term operation at elevated temperature is important. Aging is considered as one of the degradation. The effect of aging on short term property is analyzed. Material strength standard is also necessary for very thick tube sheets of forgings and small diameter thin walled seamless pipes, which are made of Mod.9Cr-1Mo steel in steam generators. This paper summarized currently available data and information on the above items, and shows path forward to the development of material strength standard for 60 years design in JSFR.

Heavy hex nuts may be selected with similar hardness to the studs to avoid stripping of nut threads for pressure containing (closure bolting) and primary load bearing bolting. However, industry specification for subsea production systems, e.g. such as API Specifications 6A, 16A, 16C, and 17D, allows the use of low strength heavy hex nuts, e.g. ASTM A194 Grade 2HM or 7M, to be applied together with high strength studs, e.g. ASTM A320 Grades L43 or L7. In the refining and chemical industries the common practice is to use ASTM A193 B7 studs with 2H nuts. Calculations and testing of nuts have been performed and capacity formulas have been established for nut structural capacities. Guidance is given on selecting nut strength/hardness to avoid stripping of nut threads. ASTM specifications give minimum quality requirements during manufacture. A brief review of standard quality requirements is given and guidance for additional requirements for high integrity fasteners in order to have equivalent quality as pressure containing forgings is given. The results from this paper may be used as background for requirements in code updates and purchaser specifications.

10 CFR 50.61 and 10 CFR 50, Appendix G require the evaluation of the unirradiated nil ductility transition reference temperature (RT NDT ) determined in accordance with ASME Section III, paragraph NB-2331. For extended operation, components outside of the traditional beltline may require evaluation of RT NDT per NB-2331. Many of these components do not have sufficient test data to determine RT NDT in accordance with NB-2331 due to different test requirements during fabrication. The US NRC has provided methods in Branch Technical Position (BTP) 5-3 of NUREG 0800 to estimate a bounding RT NDT for these components using available test data. This paper compares Branch Technical Position 5-3, paragraph 1.1(4) estimates of RT NDT using 10°F Charpy V-notch data to the ASME B&PV Section III, Paragraph NB-2331 values of RT NDT , for a sample of 35 SA-508 Cl. 2 reactor vessel shell forgings. The BTP 5-3, Paragraph 1.1(4) estimates underestimate NB-2331 values by an average of 13.8°F and by as much as 40°F. This non-conservatism is attributed to the use of 10°F Charpy V-notch data for the BTP 5-3, paragraph 1.1(4) estimate and the difficulty of using only Charpy V-notch data to estimate a parameter based on Charpy V-notch and dropweight data. If BTP 5-3, paragraph 1.1(4) estimates for RT NDT are to be used in 10 CFR 50.61 and 10 CFR 50 Appendix G calculations, the bias and measurement uncertainty of BTP 5-3, paragraph 1.1(4) should be accounted for, such that the estimate of RT NDT bounds the NB-2331 RT NDT with 2σ confidence, as shown in Regulatory Guide 1.99, Revision 2.

The latest designs of Pressure Vessels (PV), coupled with recently introduced ASME section VIII Division 3 revised code requirements, have brought about new challenges in both alloy suitability and manufacture for these safety critical components. Sheffield Forgemasters International has produced and reconditioned numerous PV forgings during the last fifty years and current demand on section sizes conveys a real challenge when trying to meet specified mechanical properties. This paper describes the technological concerns encountered when manufacturing these components and discusses the limits of what can be achieved with current practice and alloy compositions. Particular attention is given to current code testing requirements and the difficulties in obtaining an acceptable balance of strength and toughness in the mid wall of thick section large forgings.

It is important that the material used to produce high-integrity pressure vessels has homogeneous properties which are reproducible and within specification. Most heavy pressure vessels comprise large forgings derived from ingots, and are consequently affected by the chemical segregation that occurs during ingot casting. Of particular concern are the compositional variations that arise from macrosegregation, such as the channels of enriched material commonly referred to as A-segregates. By causing corresponding variations in microstructure, the segregation may be detrimental to mechanical properties. Given the scale of the pressure vessel casting, the segregation cannot be removed by practically feasible heat treatments. Here we describe an investigation on the consequences of macrosegregation on the development of microstructure in a pressure-vessel steel, SA508 Grade 3. It is demonstrated that the kinetics of transformation are sensitive to the segregation, resulting in a dramatic spatial variations in microstructure. It is likely therefore that some of the scatter in mechanical properties as observed for such pressure vessels can be attributed to macroscopic casting-induced chemical segregation.

The numerous indications recently found by UT-inspection in the reactor pressure vessel shell forgings at two Belgian nuclear power plants have raised some concerns about the effects of such indications on the vessel integrity and fitness for continued service. The UT indications have been attributed to hydrogen flaking, and preliminary estimates give a density of ∼40 indications per liter, with diameter of about 10–14 mm, oriented at a shallow ∼10° angle to the vessel inner surface. This type of high-density indications would not be characterized as geometric flaws with well defined crack-tip geometry that permits high-fidelity application of traditional fracture mechanics methods. An alternative analysis approach, with higher fidelity simulation of this type of “distributed discontinuities”, is proposed, as described in this paper. From a behavioral standpoint, the UT indications at Doel 3 and Tihange 2 represent material discontinuities whose mechanical effect can be evaluated using a damage-mechanics-based constitutive model. Previously, a special multiphase damage model was developed for cladding with zirconium hydrides, of similar morphology to the Doel 3 indications, in which the metal matrix and the hydride platelets are treated as separate material phases interacting at their interfaces with appropriate constraint conditions between them to ensure strain and stress compatibility. The hydride precipitates are represented as a brittle material and the metal matrix is modeled as a ductile elastic-plastic material. This damage model was implemented in a finite element computer program, and was validated using ring-tension and ring-compression tests of cladding specimens with various hydride morphologies. The model was able to predict specimens complete stress-strain curves and failure states with very high accuracy. The above described damage model is adapted to the high-density UT indications, morphology and distribution similar to the conditions of the Doel 3 vessel. The “hydrogen flakes” are characterized in the model as distributed damage of known orientation and volume fraction. A vessel of typical geometry and radiation-dependent mechanical properties is analyzed for various values of volume fraction of hydrogen flakes, and considering a transient loading scenario that conservatively simulates pressurized thermal shock. Interlinking of the “hydrogen flakes” and propagation of damage through the wall under the specified loading condition are part of the model’s capability of directly predicting whether or not vessel failure will occur. Thus, vessel susceptibility to failure and failure margin are judged by the degree of damage propagation through the wall.

This paper provides details of an ongoing effort to characterise the microstructure of heavy section low alloy steel forgings used in high integrity applications and correlate these data to the mechanical properties of these materials. Current industry practice is to use chemical etching and visual assessment in acceptance standards to determine nominal averages for microstructure parameters. This work uses electron microscope techniques to provide information on a variety of grain and secondary-phase particle information allowing numerical comparison of key microstructure variables to mechanical properties. For instance, the transition toughness behaviour of low alloys steels is controlled by the combination of the largest grain and particle in the material, i.e. the most potent initiator for cleavage failure. Knowledge of only the average grain size is insufficient to correlate microstructure and transition toughness performance. The programme consists of three main stages: modelling key variables in the manufacturing process to predict microstructure from thermodynamic predictions, developing quantitative microstructure data on archive materials for which mechanical property data are known to allow empirical relationships to be constructed and, a final validation exercise of a complete model by production and assessment of trial forgings.