Simulated Pressurized Water Reactor Environment Research Articles

Micro-mechanical measurement of fracture behaviour of individual grain boundaries in Ni alloy 600 exposed to a pressurized water reactor environment

A micro-mechanical technique is used to determine values of the critical stress intensity factor for fracture for grain boundaries of various orientations in Ni alloy 600 exposed to Pressurized Water Reactor (PWR) primary water at 325°C with a hydrogen partial pressure of 30kPa. Pentagonal cross-section cantilevers 5μm wide by 25μm long were milled using a focused ion beam (FIB) at individual grain boundaries in unoxidised Alloy 600 samples and in samples that had been exposed to simulated PWR environment for 4500h and for 1500h. The cantilevers were notched at the grain boundaries using FIB and tested using a nanoindenter to deflect them in bending. The critical stress intensity factor for the fractured cantilevers in samples that had been exposed for 4500h was measured to be between 0.73 and 1.82MPa(m)1/2. No intergranular fracture occurred in the samples that had been exposed for 1500h and in the unoxidised samples. No direct correlation was observed between the grain boundary misorientation angle and the critical stress intensity factor.

Effects of mixed strain rates on low cycle fatigue behaviors of austenitic stainless steels in a simulated PWR environment

Though strain rates are hardly fixed during the actual transients in nuclear power plants (NPP), most of the existing environmental fatigue ε–N data were obtained under simple loading histories with constant strain rates. In an effort to incorporate the actual loading conditions, strain rate was changed during the low cycle fatigue (LCF) test. The LCF tests were performed in a strain control mode in both simulated pressurized water reactor (PWR) environment and 310°C air to evaluate the effects of mixed strain rates on fatigue life. The test results indicated that the modified strain rate approach and average strain rate method could be applied to the fatigue evaluation considering environmental effects. The cyclic stress response showed that the trend of hardening with mixed strain rate was consistent with a negative strain rate sensitivity, which was observed in the constant strain rate test. In addition, the increase in striation spacing on the fracture surface is in good agreement with the increase of bulk dislocation density and reduction of fatigue life.

Effect of dissolved hydrogen on the corrosion behavior of chemically vapor deposited SiC in a simulated pressurized water reactor environment

The corrosion behavior of chemically vapor deposited SiC was investigated in relation to its application as fuel cladding in a pressurized water reactor (PWR). Corrosion tests were conducted in deoxygenated, pressurized water at 360°C with control of dissolved hydrogen. The dissolved hydrogen dramatically reduced the corrosion rate of SiC compared with that in water without the dissolved hydrogen. It was revealed that the dissolution of the surface oxide of SiC primarily contributed to the weight loss at the initial stage of corrosion. Further weight loss was minimal for 210 d because dissolved hydrogen effectively retarded the formation of SiO2.

Effects of dissolved oxygen on corrosion fatigue cracking of Alloy 690(TT) in pressurized water reactor environments

Alloy 690(TT) is widely used for steam generator tubes in pressurized water reactor (PWRs), where it is susceptible to corrosion fatigue. In this study, the effects of dissolved oxygen on the corrosion fatigue cracking behavior of Alloy 690(TT) in simulated PWR environments were investigated under small scale yielding conditions. Dissolved oxygen, combined with the monotonic plastic zone size (rp), had a significant influence on out-of-plane cracking of corrosion fatigue crack and environment factor (Fen).

Effects of crack tip plastic zone on corrosion fatigue cracking of alloy 690(TT) in pressurized water reactor environments

Alloy 690(TT) is widely used for steam generator tubes in pressurized water reactor (PWR), where it is susceptible to corrosion fatigue. In this study, the corrosion fatigue behavior of Alloy 690(TT) in simulated PWR environments was investigated. The microstructure of the plastic zone near the crack tip was investigated and labyrinth structures were observed. The relationship between the crack tip plastic zone and fatigue crack growth rates and the environment factor Fen was illuminated.

Intergranular Stress Corrosion Cracking growth perpendicular to fatigue pre-cracks in T–L oriented compact tension specimens in simulated Pressurized Water Reactor primary water

Intergranular Stress Corrosion Cracking (IGSCC) growth observed perpendicular to the plane of fatigue pre-cracks in T–L oriented compact tension specimens made of highly cold-rolled Type 316 stainless steel has been investigated metallurgically. One-directional cold rolling resulted in significant lamellar local strains on the plane parallel to the rolling plane. Oxide was detected along intergranular crack. It is deduced that intergranular cracking was due to selective oxidation and enhanced high diffusivity at grain boundaries coupled with lamellar local strains. It is argued that strain-induced oxidation could be a possible mechanism for IGSCC in cold-rolled stainless steels in simulated Pressurized Water Reactor environment.

Nanostructure and local properties of oxide layers grown on stainless steel in simulated pressurized water reactor environment

The morphology and local electrical resistance of duplex oxide films formed on 316L stainless steel at 325°C in simulated primary coolant of pressurized water reactor have been investigated at the nanometre scale with Conductive Atomic Force Microscopy. The electrical resistance varies over ∼1 order of magnitude for most oxide grains and over 2–3 orders of magnitude locally at grains and grain boundaries. This is rationalized in terms of local variation of the composition and thus resistivity of the mixed Fe(II)–Cr(III) barrier inner layer of the oxide films with grain boundaries of the outer layer possibly promoting Cr(III) enrichment.

Deterministic Formulation of the Effect of Stress Intensity Factor on PWSCC of Ni-Base Alloys and Weld Metals

The fundamental correlations such as crack growth rate (CGR) versus K for primary water stress corrosion cracking (PWSCC) of nickel-base alloys in simulated pressurized water reactor environments are quantified with the theoretical model based on the combination of crack tip mechanics and oxidation kinetics. Materials reliability program (MRP) proposed a CGR disposition curve in a report MRP 55 for PWSCC of thick-section Alloy 600 materials. This deterministic CGR equation has been adopted by Section XI Nonmandatory Appendix O of the ASME Boiler and Pressure Code for flaw evaluation. MRP also proposed a CGR disposition curve in a report MRP 115 for PWSCC of Alloy 82/182/132 weld metals. Stress intensity factor (K), temperature and thermal activation energy are included in both MRP 55 and MRP 115 reports. Both MRP 55 and MRP 115 are engineering-based. The results of mechanism-based modeling are compared with the screened experimental data for typical PWSCC systems of nickel-base alloys and the consistence is observed.

Cyclic Crack Growth Behavior of Reactor Pressure Vessel Steels in Light Water Reactor Environments

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533B-1 steels in simulated 550°F Boiling Water Reactor (BWR) and 550°F Pressurized Water Reactor (PWR) environments. Areas investigated over the course of the test program included the effects of loading frequency and R ratio (Kmin/Kmax) on crack growth rate as a function of the stress intensity factor (ΔK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by ΔK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The Higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were higher in a 550°F simulated BWR environment than in a 550°F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency. Further, the loading frequency at which the highest cyclic crack growth rate was observed was found to be a function of the applied ΔK level. In most cases, all cyclic crack growth rates were on or under the ASME Section XI high R water reference flaw growth line and above the Section XI air reference flaw growth line, supporting the position of these lines on the growth rate–ΔK level graph.

The Effect of Exposed Nonmetallic Inclusions on Corrosion Resistance of Type 304 Stainless Steel a Simulated Pressurized Water Reactor Environment

Abstract This paper describes tests conducted on samples of Type 304 tubing from which fuel cladding in the replacement core for the Nuclear Ship Savannah was manufactured. The major objective of the tests was to determine if exposed nonmetallic inclusions in the tubing would have a detrimental effect on the corrosion resistance of the material in the N.S. Savannah's pressurized water reactor environment. Inclusions were located by ultrasonic techniques and were exposed by metallographic methods. Two general types of inclusions were present: (1) stringers, consisting of complex Mn-Fe-Cr oxides ranging in length from 0.005 to 0.050 inch, and (2) blocks, consisting of complex Mn-Fe-Cr silicates with a maximum length of 0.010 inch. Inclusion thickness varied from less than 0.001 inch for stringers to about 0.002 inch for the block type; maximum width of both types was approximately 0.004 inch. Tests were conducted in a 5 gallon autoclave made of Type 316 stainless steel. Test solutions consisted of deoxygena...