Reactor Primary Water Research Articles

Influence of Mean Strain on Fatigue Life of Stainless Steel in Pressurized Water Reactor Water Environment

Abstract The Influence of application of the mean strain on the fatigue life was investigated for Type 316 stainless steel in the simulated pressurized water reactor (PWR) primary water environment. Low-cycle fatigue tests were conducted for a constant mean strain by controlling the strain range to be 1.2%. The applied strain rates were 0.4%/s, 0.004%/s, or 0.001%/s. The applied mean strain was 15% in nominal strain. In addition, cold worked specimens were also used for the tests without applying the mean strain. The cold working simulated the application of mean strain without an increase in surface roughness due to the application of plastic deformation. By inducing the cold working at low temperature, the influence of martensitic phase on the fatigue life was also examined. The PWR water environment reduced the fatigue life and the degree of the fatigue life reduction was consistent with the prediction model of the code issued by the Japan Society of Mechanical Engineers (JSME) and NUREG/CR-6909 Rev. 1. Increases in the maximum peak stress and stress range caused by cold working did not cause any apparent change in the fatigue life. It was revealed that the 10.5 wt% martensitic phase and the increase in the surface roughness caused by the application of 15% mean strain did not bring about further fatigue life reduction. The current JSME and NUREG/CR-6909 models were applicable for predicting the fatigue in the PWR water environment even when the mean strain or cold working was applied.

Detection of Neutrons Emitted From Reactor Primary Circuit Water by Discontinuing Flow Method

Abstract On-line activity measurement of fission products in a primary circuit water is often used for a fuel failure detection in research and power nuclear reactors. When gamma spectrometry is used for the activity measurement, high signal from 16N radionuclide and other activation products make the detection of fission products difficult. The detection of delayed neutrons emitted from several fission products is also used; however, if the detector is placed near the outlet coolant pipe, the signal from the delayed neutrons cannot be distinguished from the neutrons emitted due to 17N decay and deuterium photofission, with exception of a reactor scram condition. In this paper, a method of discontinuing the flow of primary circuit water is described. This method is based on the water flowing through a bypass on the outlet pipe to the sampling container and the flow is periodically temporarily interrupted, e.g., using 200 s + 200 s cycles. Neutrons located in the vicinity of the sampling container are continuously detected with a measuring sampling time of less than 2 s. The signal part, corresponding to the delayed neutrons, is evaluated by the signal decay analyzing during the flow interruption. The main sources of delayed neutrons suitable for this method are 137I, 87Br, and 88Br radionuclides with half-lives of 24.5 s, 55.7 s, and 16.5 s, respectively. The method was theoretically analyzed and experimentally verified in the LVR-15 research reactor.

Gauge-Strain-Controlled Air and PWR Fatigue Life Data for 304 Stainless Steel—Some Effects of Surface Finish and Hold Time

We performed environmental fatigue testing in simulated primary water reactor (PWR) primary water and reference fatigue testing in air in the framework of an international, collaborative project (INCEFA-PLUS), where the effects of mean strain and stress, hold time, strain amplitude and surface finish on fatigue life of austenitic stainless steels in light water reactor environments are being studied. Our fatigue lives obtained on machined specimens in air at 300 °C lie close to the NUREG/CR6909 mean air fatigue curve and are in line with INCEFA-PLUS air fatigue lives. Our environmental fatigue lives obtained in simulated PWR primary water at 300 °C lie relatively close to the NUREG/CR6909 mean fatigue curve; derived from the NUREG/CR6909 mean air fatigue curve and the applicable environmental correction factor (Fen). The PWR results show that (1) a polished surface finish has a slightly higher and a ground surface finish a slightly lower fatigue life than the NUREG/CR6909 prediction; (2) the ratio of polished to ground specimen life is ~1.37 at 300 °C and ~1.47 at 230 °C; (3) holds—at zero strain after a positive strain-rate—have a slightly detrimental effect on fatigue life. These results are in line with the INCEFA-PLUS PWR fatigue lives. A novel gauge-strain extensometer was deployed in order to perform a true gauge-strain-controlled fatigue test in simulated PWR primary water.

Technical Basis Part 2 for Code Case N-889: Reference Stress Corrosion Crack Growth Rate Curves for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments that were adopted as ASME B&PV Section XI Code Case N-889. As described in a technical basis Part 1 paper, over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC) and pressurized water reactor (PWR) primary water environments. Part 1 also describes the custom fitting process, quality of fit, and comparisons with related literature and data not used for fitting. This technical basis Part 2 paper describes shifting the mean models to the 75th percentile of the calibration data, simplifying to produce the N-889 curves, and comparing with previous reference curves and over 500 data points not used for developing the N-889 curves, including weld, cast, and heat-affected-zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations using both yield stress and CGR equations.

Technical Basis Part 1 for Code Case N-889: Data Selection and Modeling of Stress Corrosion Crack Growth Rates for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) models and reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments. The curves have been approved as ASME B&PV Section XI Code Case N-889, and this paper is the technical basis Part 1 for Case N-889, summarizing collection, review, and selection of calibration data, modeling, and analysis of fit. Over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC), and pressurized water reactor (PWR) primary water environments. This Part 1 paper also discusses related literature and compares with other data not used for mean model calibration. A technical basis Part 2 paper describes shifting the mean models to the 75th percentile, simplifying to produce Case N-889 reference curves, and comparing with previous reference curves and over 500 data points not used for reference curve development, including weld, cast, and heat affected zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations.

Corrosion Behavior of High-Strength Low-Alloy Steel in High-Temperature Water with Zinc and Aluminum Simultaneous Injection

Oxide films were formed on A508-3 steel in simulated pressurized water reactor (PWR) primary water at the temperature of 561 ± 1 K for 168 h with zinc and/or aluminum injection. Corrosion behaviors of oxide films were analyzed by electrochemical polarization, electrochemical impedance spectroscopy, Mott-Schottky plots, photocurrent measurement, scanning electronic microscopy, and x-ray diffraction. The results showed that zinc and aluminum simultaneous injection technology decreased the corrosion current density, increased the impedance value, made the oxide film more compact, and affected the semiconductor properties of the oxide film. The increase in zinc concentration improved the corrosion resistance to some extent. ZnAl2O4 phase, with extremely low solubility and high stability, had been detected in the oxide film; this substance changed the composition of the oxide film and affected the corrosion behavior of A508-3 steel.

Using local GND density to study SCC initiation

Strain and geometric necessary dislocations (GNDs) have been mapped with nm-resolution around grain boundaries affected by stress corrosion cracking (SCC) or intergranular oxidation with the aim of clarifying which local conditions that trigger SCC initiation of Alloy 600 in primary water reactor (PWR) water environment. Regions studied included the cracked and uncracked portion of the same SCC-affected grain boundaries and a comparable grain boundary in the as-received condition. High-resolution “on-axis” Transmission Kikuchi Diffraction (TKD) was used to generate strain and GND density based on the cross-correlation image processing method to probe shifts of specific zone axis in the TKD patterns from all regions. All cracked boundaries analyzed had local GND densities higher than 1 × 1016 m-2. Similar grain boundaries, from as-received samples had GNDs of 5 × 1014 m-2, while an intermediate level was found in the oxidized but uncracked portion of the same GB. Results, together with a discussion on the advantages and limitations of the approach, will be presented.

Influence of Deformation and Annealing on Microstructure and Corrosion Behavior of Austenitic Stainless Steel

Microstructure and corrosion behavior of warm deformed 316LN stainless steel was investigated in primary water reactor (PWR) environment by electrochemical corrosion test methods and SEM. An attempt has been made to understand the effect of warm deformation and heat treatment on the microstructure and electrochemical behavior of 316LN stainless steel. The corrosion rate is drastically increases after deformation as compare to un-deformed samples because protective oxide film is destroyed after deformation. The results indicate that the corrosion rate decrease after stress relief and solution annealing of deformed samples because heat treatment improved the repassivation capacity of deformed samples. The Ecorr shifted toward the passive direction after stress relief and solution annealing of deformed samples which indicate that a more protective passive film (Cr2O3) form on the surface of the deformed samples. It can be seen that the stress relief and solution annealing heat treated samples have less pit as compare to un-deformed samples which indicate that deformed samples have more pitting resistance as compare to un-deformed samples. Surface of the samples after corrosion test in PWR solution is investigated by SEM and it is observed that number of pits and depth of pits reduced after solution annealing and stress relief treatment of deformed sample as compare to un-deformed samples

Influence of Strain Rate and Waveshape on Environmentally-Assisted Cracking during Low-Cycle Fatigue of a 304L Austenitic Stainless Steel in a PWR Water Environment

In this paper, the low cycle fatigue resistance of a 304L austenitic stainless steel in a simulated pressurized water reactor (PWR) primary water environment has been investigated by paying a special attention to the interplay between environmentally-assisted cracking mechanisms, strain rate, and loading waveshape. More precisely, one of the prime interests of this research work is related to the consideration of complex waveshape signals that are more representative of solicitations encountered by real components. A detailed analysis of stress-strain relation, surface damage, and crack growth provides a preliminary ranking of the severity of complex, variable strain rate signals with respect to triangular, constant strain-rate signals associated with environmental effects in air or in PWR water. Furthermore, as the fatigue lives in PWR water environment are mainly controlled by crack propagation, the crack growth rates derived from striation spacing measurement and estimated from interrupted tests have been carefully examined and analyzed using the strain intensity factor range ΔKε. It is confirmed that the most severe signal with regards to fatigue life also induces the highest crack growth enhancement. Additionally two characteristic parameters, namely a threshold strain εth* and a time T*, corresponding to the duration of the effective exposure of the open cracks to PWR environment have been introduced. It is shown that the T* parameter properly accounts for the differences in environmentally-assisted growth rates as a function of waveshape.

Fatigue Life of the Strain Hardened Austenitic Stainless Steel in Simulated Pressurized Water Reactor Primary Water

Over the last 20 years or so, many studies have revealed the deleterious effect of the environment on fatigue life of austenitic stainless steels in primary water reactor (PWR) primary water. The fatigue life correlation factor, so-called Fen, which corresponds to the ratio of fatigue life in air at room temperature to that in water under reactor operating conditions, has been standardized to consider the effect on fatigue life evaluation, and the formulations are function of strain rate and temperature due to their noticeable negative effect compared with other factors (Chopra and Shack, 2007, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,” Final Report, Report No. NUREG/CR-6909, ANL-06/08; Codes for Nuclear Power Generation Facilities, 2009, "Environmental Fatigue Evaluation Method for Nuclear Power Plants," JSME S NF1-2009, The Japan Society of Mechanical Engineers, Tokyo, Japan). However, mechanism causing fatigue life reduction remains to be cleared. As one of the possible approaches to examine the underlying mechanism of environmental effect, the authors focused on the effect of plastic strain, because it could lead microstructural evolution on the material. In addition, in the case of stress corrosion cracking (SCC), it is well known that the strain-hardening prior to exposure to the primary water can lead to remarkable increase of the susceptibility to cracking (Vaillant et al., 2009, “Stress Corrosion Cracking Propagation of Cold-Worked Austenitic Stainless Steels in PWR Environment,” 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems; Couvant et al., 2009, "Development of Understanding of the Interaction Between Localized Deformation and SCC of Austenitic Stainless Steels Exposed to Primary PWR Environment," 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems). However, its effect on fatigue life has not necessarily been cleared yet. The main effort in this study addressed the effect of the prior strain-hardening on low cycle fatigue life in the primary water. A plate of 304LSS was strain hardened by cold rolling or tension prior to fatigue testing. The tests were performed under axial strain control at 300 °C in primary water including B/Li and hydrogen, and in air. The effect on environmental fatigue life was investigated through a comparison of Fen in experiments and in regulations, and also the effect on the fatigue limit defined at 106 cycles was discussed.

Influence of localized plasticity on oxidation behaviour of austenitic stainless steels under primary water reactor

The sensitivity of precipitation-strengthened A286 austenitic stainless steel to stress corrosion cracking was studied by means of slow-strain-rate tests. First, alloy cold working by low cycle fatigue (LCF) was investigated. Fatigue tests under plastic strain control were performed at different strain levels (Δεp/2=0.2%, 0.5%, 0.8% and 2%) to establish correlations between stress softening and the deformation microstructure resulting from the LCF tests. Deformed microstructures were identified through TEM investigations. The interaction between oxidation and localized deformation bands was also studied and it resulted that localized deformation bands are not preferential oxide growth channels. The pre-cycling of the alloy did not modify its oxidation behaviour. However, intergranular oxidation in the subsurface under the oxide layer formed after exposure to PWR primary water was shown.

Primary water stress corrosion cracking in alloy 600 and intergranular crack growth rate

The effects of the stress intensity factor and dissolved hydrogen on the crack growth rate in primary water reactor environments was investigated with alloy 600 with continuous carbide films at grain boundaries. A primary water stress corrosion cracking of intergranular fracture mode was observed. The intergranular crack growth rate decreased with a decreasing stress intensity factor, and the dissolved hydrogen content influenced the crack growth rate. Such results are consistent with those reported previously.

Separation of Microstructural and Microchemical Effects in Irradiation Assisted Stress Corrosion Cracking using Post-irradiation Annealing

Post-irradiation annealing (PIA) was conducted in order to clarify the role of microstructural and microchemical effects on irradiation assisted stress corrosion cracking (IASCC) susceptibility in simulated pressurized water reactor (PWR) primary water. Microstructures, hardening, radiation-induced segregation and IASCC susceptibility were examined in cold-worked SUS 316 stainless steels irradiated to 25 dpa in a PWR after annealing at 673–823 K for 1h. IASCC susceptibility, microstructures and hardening recovered as the annealing temperature increased whereas the grain boundary segregation of Cr and Ni remained almost unchanged. The results suggested that the change in IASCC susceptibility due to annealing is not attributed to the change in grain boundary segregation but to the change in micro-structures and hardening. The fact that a smaller recovery of radiation hardening caused a larger IASCC susceptibility suggested that a threshold hardening level exists for the occurrence of IASCC.

Effects of Dissolved Hydrogen on the Primary Water Stress Corrosion Cracking Behavior of Alloy 600 at 325 °C

Primary water stress corrosion cracking (PWSCC) of Alloy 600 has been a great concern to the nuclear power industry. Reliable PWSCC growth rate data, especially at temperatures in the range of 290-330°C, of the alloy are required in order to evaluate the lifetime of power plant components. In this study, three tests were carried out in simulated pressurized water reactor (PWR) primary water at 325°C at different dissolved hydrogen (DH) concentrations using standard one-inch compact tension (1T-CT) specimens. The initiation and growth of cracks as well as insights into the different PWSCC mechanisms proposed in the literature were discussed. The experimental results show that the detrimental effects of hydrogen on crack initiation and growth reached a maximum at a certain level of DH in water. The experimental results were explained in terms of changes in the stability of the surface oxide films under different DH levels. The experimental results also support the assumption that hydrogen absorption as a result of cathodic reactions within the metal plays a fundamental role in PWSCC.

Environmentally Assisted Cracking of Alloy X-750 in Simulated PWR Coolant

Abstract Environmentally assisted cracking (EAC) of different heat treatments of alloy X-750 has been studied in simulated pressurized water reactor (PWR) primary water from 50 to 340°C. The alloy, with single- (704°C/20 h) and two-stage (885°C/24 h + 704°C/20 h) aging treatments, was increasingly susceptible to EAC with decreasing temperature below ∼250°C. This susceptibility was due to hydrogen embrittlement, as indicated by the sensitivity to the hydrogen content of the water and behavior in hydrogen gas. At higher testing temperatures, a second mechanism, possibly anodic dissolution, was dominant for alloys with either a lower solution heat treatment temperature (980°C/1 h) or two-stage aging treatment. This second mechanism was suppressed by a higher temperature solution treatment (∼1100°C/1 h) and a single-stage aging treatment, which improved alloy performance.

Microstructure and stress corrosion resistance of alloys X750, 718, and A286 in light water reactor environments

The results of microstructural characterization and stress corrosion cracking tests are presented for several heat treatment conditions of the age-hardenable Alloys X750, 718, and A286. Resistance to intergranular stress corrosion cracking (IGSCC) was evaluated using slow strain rate tests in pressurized water reactor (PWR) primary water, microstructural etch, and rising-load screening tests. The results of these tests indicate greater IGSCC resistance in the PWR primary water environment for X750 microstructures with heavier grain boundary coverage by Cr23C6, regardless of grain boundary chromium depletion. Alloys 718 and A 286 were not highly SCC susceptible according to these screening tests, although they are not immune to IGSCC in reactor environments. Results are compared with those of a related program, including crack initiation and crack growth tests in pressurized and boiling water reactor (BWR) environments.

Stress Corrosion of Age-Hardenable NiFeCr Alloys

Abstract Stress corrosion cracking (SCC) tests have been conducted on the age-hardenable NiFeCr Alloys X-750, 718, and A-286 in flowing pressurized water reactor (PWR) primary coolant water at 360 C. The effect of Zr additions to Alloy X-750 and the effect of heat treatment on all of the alloys were studied. The test specimens were U-bends and precracked 1/2 T-CT fracture mechanics specimens. The total exposure was 6000 h, with inspections at 500, 1500, and 3000 h. No initiation was observed on the U-bend specimens, but crack growth was observed on all of the precracked samples. The cracking was intergranular on all of the specimens, with only a few grains of transgranular cracking occurring on one. Zr additions to Alloy X-750 had no significant effect on SCC growth. The 1107 C solution anneal plus the overaged conditions provided the greatest resistance to SCC growth. Alloy 718 exhibited the least resistance.