Effect of Hydrogenated Water on Fatigue Crack Growth for Austenitic Stainless Steel in Simulated BWR Water Environment

Fatigue crack growth rates (FCGR) of sensitized austenitic stainless steel (SS) were measured in simulated BWR water at 288 °C using compact tension specimens under different cyclic loading modes, including saw-tooth, trapezoidal and constant loading pattern. This study tested sensitized SS in normal water chemistry (NWC) and hydrogen water chemistry (HWC) respectively, and attempted to clarify the effect of low electrochemical corrosion potential on the FCGR of sensitized stainless steel. Significant environment effects on FCGR of sensitized stainless steel were observed in both water chemistries when compared with air fatigue curve. The pronounced suppression effect of HWC on crack growth in statically sustained load was not observed in cyclic loading condition. ASME curve doesn’t seem to be conservative and could not bound all the FCGR data tested in this study. In contrast, all of the measured FCGR data were bound by the JSME disposition curve. PLEDGE model proposed by General Electric reasonably predicted the FCGR of sensitized SS in NWC, but underestimated the FCGR in HWC. ANL’s superposition model successfully estimated the FCGR measured in both water chemistries. The fractography exhibited transgranular fracture mode during the crack initiation and growth stage. No differences in the appearance of fracture surface were observed in HWC and NWC. Only in very high DO environments, the sensitized 304 SS exhibited the mixed mode of intergranular and transgranular during growth stage.

Fatigue Crack Propagation Tests on 304 Stainless Steel in High Temperature Water: Accelerated Cracking Rates and Transition to Lower Rates

Previous fatigue crack propagation (FCP) tests on a single heat of 304 stainless steel (304 SS) specimens showed a strong acceleration of rates in high temperature water with 40–60 cc H2/kg H2O at 288°C, with rates up to 20X the air rates. The accelerated rates were observed under fully reversed conditions (R = −1) (Wire and Mills, 2001) and high stress ratios (R = 0.7 and 0.83) (Evans and Wire, 2001). In this study, a second heat of 304 SS has been tested at 243°C and 288°C and lower positive stress ratios (R = 0.3, 0.5). The second heat showed the large acceleration of rates at 288°C observed previously. Rates were up to two times lower at 243°C, but were still 7–8X the air rates. A time-based correlation successfully correlates the accelerated rates observed, and is nearly identical to fits of literature data in hydrogen water chemistry (HWC), which has hydrogen added at a lower level of about 1 cc/kg H2O. The accelerated rates on the second heat were not stable under two different test conditions. In contrast to the first heat, the second heat showed a reduction in environmental enhancement at long rise times, accompanied by a change in fracture mode. Addition of a constant load hold time of 1200 s between cycles also caused a marked reduction in crack propagation rates in both heats, with reduction to nearly air rates in the second heat. The differing rise time effects between the two heats could be rationalized by time-dependent deformation. More hold time testing is required to define the material and loading conditions which lead to reduced rates.

What’s New in Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components

The nuclear industry is a pretty dynamic industry, in that it is always on the move, changing every time we turn around. For that very reason, there is a need to keep up with the industry by providing changes to American Society of Mechanical Engineering Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” There have been many changes over the last three years. This paper addresses a few of those, but gives a feel for the number of changes from the 2000 Addenda to the 2003 Addenda, there have been a total of approximately 56 changes. Of those changes, 11 were in the repair/replacement requirements, 19 in the inspection requirements, 4 in the evaluation requirements, 18 in the nondestructive examination requirements, and 4 in the administrative requirements. The paper classifies the changes as “Technically Significant,” “Significant,” “Non-Significant,” or “Editorial.” The paper addresses only a few of those changes that were “Technically Significant.” The paper also includes some of the activities that the ASME Section XI Subcommittee is currently working on.

Fatigue Crack Growth Curve for Austenitic Stainless Steels in PWR Environment

Although reference fatigue crack growth curves for austenitic stainless steels in air environments and boiling water reactor (BWR) environments were prescribed in JSME S NA1-2002, similar curves for pressurized water reactors (PWR) were not prescribed. In order to propose the reference curve in PWR environment, fatigue tests of austenitic stainless steels in simulated PWR primary water environment were carried out. According to the procedure to determine the reference fatigue crack growth curve of BWR, which of PWR is proposed. The reference fatigue crack growth curve in PWR environment have been determines as a function of stress intensity factor range, Temperature, load rising time and stress ratio.

Experimental Study for Proximity Rule of Multiple Flaws for Ductile Fracture

The proximity rule of multiple flaws in ASME B&PV Code Section XI 2003 addenda was mainly determined by the evaluation of stress intensity factors from the viewpoint of brittle fracture. Since the austenitic steel and carbon steel for class 1 piping shows a ductile manner in fracture, a new proximity rule for ductile fracture is required. To understand the fracture behavior of multiple flaws, tensile tests, using flat plate specimens made of Type 304SS with twin flaws, were conducted. When the shapes of twin flaws were semi-circular with aspect ratio, a/l = 0.5, the effect of the space of multiple flaws on maximum load is clear. On the other hand, the effect of flaw spacing on maximum load was insignificant for flat multiple flaws with 0.167 in a/l. The effect of space of multiple flaws, aspect ratio of multiple flaws on ductile fracture pattern was discussed. Finally, the proximity rule for plastic collapse was proposed in this paper.

What’s New in Section I [2001 Through 2003]

It has been a three years since I last prepared a summary of significant revisions to the Section I of the ASME Boiler and Pressure Vessel Code. In addition to the routine maintenance and upgrade revisions that occur in each annual addenda to the Code, there are several major activities under way that will have a significant impact on Section I in the future. One of these is a contemplated new Part covering Heat Recovery Steam Generators. The second is the publication of the 2004 Edition of the Boiler Pressure Vessel (B&PV) Code in SI units.

Study on Flaw Acceptance Standard of ASME Code Sec. XI Based on Failure Probability

Based on the failure probability, the flaw acceptance standard of ASME Code Sec. XI is examined with some concerns weather the failure probability is uniform for flaws with various aspect ratios and failure frequencies are small enough. In this paper, the results of preliminary case studies are described on the failure probability of reactor pressure vessels (RPVs) with a surface flaw specified in Sec. XI. PFM code PASCAL was used for case studies. A PTS (Pressurized Thermal Shock) transient prescribed by NRC/EPRI PTS Benchmark Study was used as an applied load. Analysis results showed that the conditional failure probability of a RPV with an initial flaw of acceptable depth depends on the aspect ratio. In the case flaw shapes are close to semi-circular, the failure probability are higher than that of the cases aspect ration are less than 0.6 by one order of magnitude due to the difference of fracture behavior at the surface point. A case study for determining the acceptable flaws based on failure probability was also carried out.

A Parametric Study of ASME B16.5 Flanges Which Has Experienced Flange Face Corrosion

Weld neck flanges on piping systems are susceptible to flange face corrosion when they are exposed to corrosive environments. This paper examines the maximum amount of corrosion a weld neck flange face could tolerate without loosing structural integrity and hence the flange is fit for service. A parametric study using finite element method was used to examine the entire range of weld neck flanges listed in ASME B16.5 Code, Pipe Flanges and Flanged Fittings. From the study, a number of tables were developed limiting the amount of corrosion for the various classes and sizes of flanges.

Design Issues and Considerations During Reactor Vessel Head Replacement, Steam Generator Replacement, and Steam Generator Snubber Elimination at Nuclear Power Plants

One of the key maintenance activities in a nuclear power plant is the replacement of major components in the Nuclear Steam Supply System. In order to achieve significant operational improvements, the replacement components are not an exact replacement of the existing components. The replacement of components in the nuclear steam supply system in many Pressurized Water Reactor plants may include steam generators, replacement of reactor vessel heads with integrated head assemblies, and elimination of steam generator snubbers. The replacement components may not be supplied and/or designed by the original supplier. The changes in the components have to be compared to a plant’s current design and licensing bases and regulatory commitments. The qualification of these components involves non-linear, Nuclear Class 1 analyses, where portions of the configuration and analyses are proprietary, and there is a coupling of the response between the containment structure and the components. Ultimately, the qualification of the reactor coolant system and reactor vessel internals must be demonstrated, not just the qualification of the replacement components. A key element for the successful completion of these component replacements is the method by which the design and licensing bases is maintained and the work of the various groups involved in the design coordinated. This paper outlines how in a typical two unit PWR plant, major component replacements can impact original design bases and issues that should be considered in creating successful design and configuration documents. Design interface issues, configuration combinations, and coordination requirements are identified.

Background and Technical Basis for the Evaluation of Flaws in PWR Reactor Vessel Head Penetration Regions

Acceptance criteria have been developed for indications found during inspection of reactor vessel in upper head penetrations. These criteria were originally developed for inside surface flaws, as part of an industry program coordinated by NUMARC (now NEI) in 1992. These criteria were not inserted into Section XI at the time, because inspections were not required for these regions. In developing the enclosed acceptance criteria, the approach used by the industry group was similar to that used in other portions of Section XI, in that an industry consensus was reached using input from the operating utility technical staff, each of the three PWR vendors, and representatives of the NRC staff. The criteria developed are applicable to all PWR plant designs. The discovery of leaks at Oconee, ANO-1, and several other plants, have led to the imposition of inspection requirements for head penetration regions, and therefore the need to develop criteria for indications in all portions of the tubes. This would include indications on the inside diameter of the tube, as well as on the outside diameter of the tube below the attachment weld, and flaws in the attachment weld itself. The criteria presented herein are limits on flaw sizes which are acceptable. The criteria are to be applied to inspection results. It should be noted that determination of the period of future service during which the criteria are satisfied is plant-specific and dependent on flaw geometry and loading conditions. It has been previously demonstrated by each of the owners groups that the penetrations are very tolerant of flaws. It was concluded that complete fracture of the penetration would not occur unless very large through-wall flaws were present; therefore, protection against leakage during service is the priority. The approach used here is more conservative than that used in Section XI applications where the acceptable flaw size is calculated by putting a margin on the critical flaw size. In this case, the critical flaw size is far too large to allow a practical application of this approach, so protection against leakage is the key element used to define the acceptance criteria. Also, the use of flaw acceptance standards tables is not allowed for this region, for penetrations which are susceptible to stress corrosion cracking. The acceptance criteria apply to all flaw types regardless of orientation and shape. The same approach is used by Section XI, where flaws are characterized according to established rules and their future predicted size is then compared with the acceptance criteria.