Keywords

1 Introduction

There are pressurized water reactor (PWR), boiling water reactor, heavy water reactor and other types of commercial nuclear power plants. PWR is the most widely used because of its mature technology and rich operation experience, accounting for more than half of the operating nuclear power plants in the world. Prolonged neutron irradiation leads to the changes of mechanical properties due to irradiation induced hardening effect, the changed of local chemical compositions due to radiation-induced segregation effect, the increase of physical defects such as dislocation loops, and the water decomposition due to radiolysis, and finally affects the SCC in high temperature water.

IASCC is generally divided into three stages: crack initiation, crack propagation and instable fracture [1]. At the initial stage of crack initiation, the surface of sensitive materials begins to produce microcracks under the coupling effect of environmental and mechanical factors, and these microcracks are invisible under the light microscope. Over time, the microcracks merge with each other to form an initial crack with a length of 10 μm. In the crack growth stage, the crack expands at a certain rate, which is affected by environmental and mechanical factors. The laboratory usually studies the process by introducing the stress intensity factor K at the crack tip. In the instable fracture stage, the crack expands rapidly until the material fracture.

It is well known that SCC of materials in high temperature and high pressure water depends on three factors: materials, environment and relatively high stress is shown in Fig. 1. The key parameters affecting IASCC mainly include the material itself (such as microstructure, microchemistry and yield strength, etc.) and environmental parameters (such as hydrochemistry, irradiation temperature and irradiation dose, etc.). The key parameters such as irradiation temperature and irradiation dose have great influence on the crack growth rate of austenitic stainless steel IASCC. There are also many studies on the influence of material factors such as microstructure, microchemistry and yield strength on IASCC sensitivity of austenitic stainless steel.

Fig. 1.
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The main influencing factors of IASCC

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The study of IASCC behavior of austenitic stainless steel needs to pay attention to three stages: crack initiation, crack propagation and instable fracture. The crack propagation stage is the most important. Stress corrosion cracking promoted by irradiation is affected by material itself, medium environment, irradiation temperature and irradiation dose, etc. The research process is complicated and uncertain. The relationship between CGR and K is the key parameter in IASCC process of austenitic stainless steel.

SCC or IASCC is used for the study of materials. There are many types of specimens, such as compact tension (CT) specimen [2], four-point bending specimen [3], round compact tension (RCT) specimen [4] and so on. In this paper, the advantages and disadvantages of sample types are not discussed too much, and only the experimental results are concerned.

2 Compiling IASCC Data of Austenitic Stainless Steels in Simulated PWR Primary Water

Austenitic stainless steels can be divided into many types, among which 304 and 316 based austenitic stainless steel is widely studied, so here only around 304 and 316 based austenitic stainless steel is discussed.

The stress corrosion cracking experiment promoted by neutron irradiation needs special equipment, and the experimental difficulty exists objectively due to the limitation of equipment and other conditions. Radiation doses can range from a few percent of dPa to more than 100 dPa. Experimental data on crack growth rate of 0.06–47.0 dPa austenitic stainless steel were collected in this paper.

The CGR data of irradiated austenitic stainless steel in simulated BWR or PWR environments were obtained from CT samples, RCT samples or four-point bending samples. In the experiment, factors such as irradiation temperature, material type and water environment should be controlled, and the irradiation dose should be reasonably controlled. The CGR data of the sample should be matched with K one by one, and the relation data between CGR and K should be obtained.

Based on 304 and 316 austenitic stainless steel, the materials were irradiated in the light water reactor environment, and the data of researchers were summarized to obtain Fig. 2 [4,5,6,6,7,8,9,10,11,12].

Figure 2 is a summary of the relationship between CGR and K of irradiated austenitic stainless steel in a pressurized water reactor environment sorted by irradiation dose. Due to the limitation of experimental data conditions of many researchers, it is difficult to unify the irradiation temperature, which is controlled at 288–340 ℃. The influence of irradiation temperature on CGR is very complex and limited by space, so the irradiation temperature will not be discussed too much in this paper. Instead, data analysis and problems will be found dialectically.

The parameter stress intensity factor K is introduced here, which essentially reflects the mechanical changes in the microstructure. Microcracks exist in the microstructure of irradiated austenitic stainless steel. If two cracks have the same strain and stress in a region near the crack tip, then they have the same K. Since the crack tip scale is small, K represents the stress and strain at the crack tip.

As shown in Fig. 2, as a whole from the reference curve, CGR generally shows an increasing trend with the increase of K. However, we found that after data integration, each data was not strictly linear and was greatly affected by the radiation dose.

The CGRs of SCC of unirradiated austenitic stainless steel is lower than that of IASCC of irradiated austenitic stainless steel under the same conditions, and the difference may be several times. Taking the CGRs data of austenitic stainless steel under 3 dpa irradiation dose as an example, the CGR of unirradiated austenitic stainless steel is about 3x10–10 m/s [2,3,4,5] while the CGR of irradiated austenitic stainless steel can reach 9x10–10 m/s or 1x10–9 m/s. The difference is about three times. This indicates that neutron irradiation can promote the crack growth rate (CGR) of austenitic stainless steel IASCC to a certain extent. The effect of irradiation dose on CGR of austenitic stainless steel IASCC in PWR is complicated. From the figure, we can roughly divide the radiation dose into low dose, medium dose and high dose. Low radiation dose mainly refers to 0–9 dpa. Under this condition, the change of CGR is not obvious compared with that without radiation. K is mainly between 12 and 20 MPa m0.5. Based on literature review and PWR experience, 3 dpa is the threshold of material irradiation. In the case of 0–3 dpa irradiation dose, the CGR of the material will not change with the increase of the irradiation dose, but still maintain the original growth rate. Under the irradiation dose of 3–9 dpa, irradiation promoted the CGR of austenitic stainless steel. For example, the CGR of 3 dpa was higher than that of 6.3 dpa and 8.0 dpa. K remained in a certain range, while CGR changed abnormally. It is speculated that the CGR is affected by the changes of microstructure defects, dislocation loops and radiation dose.

Fig. 2.
figure 2

Relationship between CGR and K of austenitic stainless steel at 0.6–47.0 dpa in pressurized water reactor [4,5,6,7,8,9,10,11,12]

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The radiation dose here refers to the radiation dose of approximately 11.4–25.0 dpa. In the case of medium irradiation dose, CGR and K maintain a positive correlation. At 17.1 and 18.7 dpa, the data are mainly concentrated at the lower left of the figure near the reference curve. At 21.0, 12.9 and 14.0 dpa, the data mainly concentrated in the center of the graph near the curve. Although the data at 11.4, 15.0, and 25.0 dpa are mainly distributed in the upper right of the graph, they generally follow the trend of increasing CGR as K increases. In these data, the phenomenon of 25.0 dpa is worth separate discussion and will be carried out in the future.

High radiation dose refers to radiation dose above approximately 25.0 dpa. As can be seen from the figure, although the CGR of austenitic stainless steel under high irradiation dose is still higher than that of non-irradiated stainless steel, compared with medium irradiation dose, the CGR of austenitic stainless steel is abnormally reduced, mainly between 1x10–11 m/s and 1x10–10 m/s. These results indicate that high irradiation dose is not conducive to the acceleration of the growth rate of austenitic stainless steel IASCC in PWR, but has a certain inhibitory effect.

The relationship between CGR and K of austenitic stainless steel in pressurized water reactor remains relevant, and CGR increases with K in most cases. However, with the increase of irradiation dose, CGR and K of austenitic stainless steel do not increase simultaneously, but atrophy occurs. It may be caused by the excessive damage to the material caused by neutron irradiation.

It is worth noting that although the CGR and K relationship data of irradiated austenitic stainless steel were partially dispersed, CGR and K still showed a positive correlation. It should be pointed out that the 316 based austenitic stainless steel material under the neutron irradiation dose of 25 dpa in Fig. 2 has abnormal conditions. At lower K, that is, when K is less than or equal to 5 MPa m0.5, the CGR of the material is more than 9x10–11 m/s. The appearance of abnormal data points in Fig. 2 caught our attention and thought.

3 Effect of Irradiation Dose on SCC Growth Rates

The CGR of austenitic stainless steel in high temperature water is affected not only by stress intensity factor K, but also by neutron irradiation dose, which is indispensable in the process of stress corrosion cracking promoted by irradiation. In order to study the effect of neutron irradiation dose on CGR of austenitic stainless steel in high temperature water, temperature and K must be controlled within a certain range to make the data reasonable and reliable.

Firstly, the temperature was selected between 335 and 340 ℃, which could remove the interference of temperature factors on crack growth rate. Through Fig. 2, we summarized the CGR data of 0.06, 3.0, 8.0, 9.0, 11.4, 12.9, 14.0, 15.0, 17.0, 18.7, 21.0, 32.9, 37.8, 39.0, 47.0 dpa irradiation dose, as shown in Fig. 3 [4,5,6,7,8,9, 11].

Fig. 3.
figure 3

CGR and K of materials with different neutron irradiation doses at 335–340 ℃ [4,5,6,7,8,9, 11]

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Here we classify K as 5–10 MPa m0.5, 10–15 MPa m0.5, 15–20 MPa m0.5 and 20–27 MPa m0.5. The CGR and K data of austenitic stainless steel at 335–340 ℃ were processed to obtain the CGR and neutron irradiation dose of austenitic stainless steel under different normalized K conditions.

Fig. 4.
figure 4

CGR and neutron irradiation dose of austenitic stainless steel under normalized K 5–10 MPa m0.5 [5,6,7,8,9, 11]

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When K is normalized to 5–10 MPa m0.5, as shown in Fig. 4 [5,6,7,8,9, 11], the CGR of the material is the highest at 11.4 dpa, close to 1x10–8 m/s. When K is 5–10 MPa m0.5, CGR of all irradiation dose data is greater than 9x10–12 m/s. At 17.1 dpa and 18.7 dpa, the CGR values of the materials are close to each other, basically within the range of 10–11 m/s and 10–9 m/s. At 21.0 dpa, the CGR of the material increased slightly to about 10–9 m/s. CGR at 47 dpa was basically within the data range of 17.1 dpa and 18.7 dpa irradiation dose.

Fig. 5.
figure 5

CGR and neutron irradiation dose of materials under normalized K 10–15 MPa m0.5 [4,5,6,7,8,9, 11]

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As shown in Fig. 5 [4,5,6,7,8,9, 11], when K is in the range of 10–15 MPa m0.5, CGR of all data in the figure is greater than 4x10–11 m/s. CGR data under 3.0 dpa were higher than 10–10 m/s. The CGR data range of 12.9 dpa and 14.0 dpa irradiation doses were similar. The CGR at 15.0 dpa was two orders of magnitude higher than that at 12.9 dpa and 14.0 dpa. The CGR data at 21.0 dpa irradiation dose were about 1 order of magnitude higher than those at 12.9 dpaand 14.0 dpa irradiation dose. The CGR data of 21.0 dpa irradiation dose was about 0.1 times that of 15.0 dpa irradiation dose. When the neutron irradiation dose increased from 32.9 dpa to 37.8 dpa and from 37.8 dpa to 47.0 dpa, the CGR data at these three irradiation doses were close to 10–10 m/s (between 5x10–11 m/s and 1x10–10 m/s). The CGR data at these three irradiation doses are not sensitive to the neutron irradiation dose.

Fig. 6.
figure 6

CGR and neutron irradiation dose of materials under normalized K 15–20 MPa m0.5 [4,5,6,7,8,9, 11]

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CGR (3.0 dpa) < CGR(21.0 dpa) < CGR(15.0 dpa) > CGR(12.9 dpa, 14.0 dpa, 32.9 dpa, 37.8 dpa and 47.0 dpa). The peak value of CGR data with neutron irradiation dose of 15 dpa in Fig. 5 has aroused our concern. As for the reason of the peak value? What causes the spike remains to be studied and solved.

As shown in Fig. 6 [4,5,6,7,8,9, 11], when K is 15–20 MPa m0.5, CGR at 0.06 dpa, a very low irradiation dose, is very low, about 2x10–12 m/s. However, when the irradiation dose reached 3.0 dpa, the CGR data directly reached nearly 10–9, with a significant increase. When the irradiation dose was higher than 8.0 dpa, the CGR of the material increased significantly, and the CGR was greater than 2x10–11 m/s. The CGR data range of 11.4 dpa was close to that of 15.0 dpa, up to about 10–8 m/s. CGR data at 12.9 dpa and 14.0 dpa are close in range, but are still two orders of magnitude lower than CGR data at 11.4 dpa and 15.0 dpa. The CGR data at 21.0 dpa was between 12.9 dpa and 14.0 dpa and 11.4 dpa and 15.0 dpa, with a value of about 10–9 m/s.

On the whole, it shows that when the neutron irradiation dose is greater than or equal to 15 dpa, the crack growth rate decreases with the increase of the dose. At the same time, the irradiation dose of 11.4 dpa and 15.0 dpa showed two similar peak values. Similar to the situation where K is 10–15 m/s as shown in Fig. 5, the phenomenon of peak value arouses concern. The two can be compared and further studied.

Fig. 7.
figure 7

CGR and neutron irradiation dose of materials under normalized K 20–27 MPa m0.5 [5, 6, 8]

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As shown in Fig. 7, when K is in the range of 20–27 MPa m0.5, the range of CGR data fluctuates greatly. The main reason is that the CGR data of 0.06 dpa is small and that of 11.4 dpa is large. When the irradiation dose increased from 0.06 dpa to 3.0 dpa, the CGR increased by about 2 steps. From 3.0 dpa to 8.0 dpa, its CGR decreased by more than one order of magnitude, but it was still about four times the CGR data at 0.06 dpa. CGR data at 11.4 dpa fluctuated from 4x10–10 m/s to nearly 10–7 m/s, with a fluctuation range of more than 2 orders of magnitude. Compared with the CGR data at 8.0 dpa, the CGR data at 39.0 dpa was 1.5 or 1.75 times higher, and the CGR data at 47.0 dpa was 2.5 or 0.75 times higher.

Among them, CGR data at 11.4 dpa fluctuated from 4x10–10 m/s to nearly 10–7 m/s, with a fluctuation range of more than 2 orders of magnitude. In the same range of K, why is there such a high CGR data, and why is there such a large fluctuation range of CGR data in the same radiation dose? These questions remain to be explored.

Based on the above data, we propose a method to judge the phase of CGR data varying with neutron irradiation dose.

$$\begin{gathered} \frac{{\delta_{CGR} }}{{\delta_{Dose} }} > {\text{0: type I,}} \hfill \\ \frac{{\delta_{CGR} }}{{\delta_{Dose} }} = {\text{0: type II}} \hfill \\ \frac{{\delta_{CGR} }}{{\delta_{Dose} }} < {\text{0: type III}} \hfill \\ \end{gathered}$$

Each type is divided into three categories: type I for increasing, type II for nearly not changing, and type III for decreasing of crack growth rate with increasing dose. There are −S for strongly, −M for moderately, and −W for weakly. As shown in Table 1, the change rate of CGR and irradiation dose in different K ranges is shown.

Table 1. Classification of CGR and radiation dose changes in different K ranges
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4 Verification of Effect of Irradiation Doses on SCC Growth Rate

Bosch [13] et al. recently studied the correlation between neutron irradiation and mechanical properties of 316 material cold worked processed in pressurized water reactor environment, which provided reference value for our paper. He studied the relationship between neutron irradiation and stress-strain, tensile strength, yield strength and other mechanical properties, and we only take the yield strength related content.

Fig. 8.
figure 8

Variation of yield strength of cold worked 316 material with neutron irradiation dose in pressurized water reactor [13]

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Figure 8 shows the change of yield strength of 316CW material in PWR environment with neutron irradiation dose under the irradiation temperature of 320–340 ℃ and the test temperature of yield strength of 300–320 ℃.

Fig. 9.
figure 9

CGR and K of 316CW material irradiated at 25.0 dpa, CGR of the material at different yield strengths when K is 30 MPa m0.5 [14,15,16,17,18].

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In Fig. 9 we can see the CGR distribution of 316CW material at 25 dpa. Through the study of Terachi [14] et al., Castano [15] et al., Shoji [16] et al., Toloczko [17] et al. and Donghai Du [18] et al., we can find the CGR situation of 316CW materials without irradiation under different yield strengths of K is 30 MPa m0.5. It can be clearly seen that the CGR of about 800 MPa yield strength is between 10–10 m/s and 10–9 m/s, which is nearly 2 orders of magnitude lower than the CGR after irradiation at the same K. CGR with yield strengths of 500 MPa and more than 200 MPa is between 10–11 m/s and 10–10 m/s, which is nearly 3 orders of magnitude lower than CGR after irradiation.

We can obviously score that the CGR of the cold worked 316 material after neutron precipitation is much higher than that of the cold worked 316 material without neutron irradiation. This shows that neutron irradiation has a great influence on the change of material CGR. In addition, the microstructure defects and dislocations of materials may be increased due to the influence of neutron irradiation on the microstructure and mechanical properties of materials.

5 Conclusions

The CGR and K relationship data of 304 and 316 austenitic stainless steels under different neutron irradiation doses in high temperature water were summarized. The effect of irradiation dose on the crack propagation rate of irradiated SCC was analyzed and verified:

The CGR and K of austenitic stainless steel under high temperature water environment show positive correlation. The CGR of unirradiated CGR was lower than that of irradiated CGR under the same condition, and the difference was several times. There is a certain degree of dispersion in the relation data between CGR and K.

In the range of different K, the relationship between CGR data and irradiation dose of austenitic stainless steel is different. Peak data points of CGR appear at some irradiation doses, and the specific reasons need to be studied. We propose a classification method for CGR and radiation dose variation trend and apply it to the data presented in this paper.

Finally, the significant effect of neutron irradiation on CGR of austenitic stainless steel was verified based on yield strength.