KR20100104928A - Method of preventing initiation of primary water stress corrosion cracking of ni-base alloy for nuclear power plant - Google Patents

Abstract

PURPOSE: A method of preventing the initiation of primary water stress corrosion cracking of Ni-base alloy structural components in a nuclear power plant is provided to prevent additional contraction because the regularization of Ni-base alloy is completed before exposure to the operating environment of the nuclear power plant. CONSTITUTION: A method of preventing the initiation of primary water stress corrosion cracking of Ni-base alloy structural components in a nuclear power plant comprises a step of maintaining Ni-base alloy at 400~520°C for 0.5~200 hours for regularization, wherein the Ni-base alloy comprises Alloy 600, Alloy 690, Weld 182, Weld 82, Weld 152, Weld 52, and Weld 52M.

Description

Method of Preventing Initiation of Primary Water Stress Corrosion Cracking of Ni-base Alloy for Nuclear Power Plant

The present invention relates to a method for preventing the initiation of primary water stress corrosion cracking (PWSCC) of a nickel-base alloy constituting a nuclear reactor and a nickel-base alloy which is a welding material connecting the parts. In more detail, the nickel base alloys of Alloy 600 and its weld metal Weld 182/82/132, Alloy 690 and its weld metal Weld 152/52 / 52M, which are used as part of nuclear reactors, are to be subjected to After completion of the regularization by heat treatment, the primary structure of the nickel-base alloy nuclear power plant structure which is exposed to the environment of the first nuclear power plant to prevent PWSCC from occurring due to the shrinkage of additional crystals due to regularization in the normal operation of the nuclear reactor. It relates to a method for preventing systematic stress corrosion cracking initiation.

Alloy 600 is used in many structural materials, such as reactor steam generator tubes, CRDM (control rod driving mechanism) nozzles, and bottom mounted instrumentation (BMI) nozzles. Weld is used for welding other metals, including Alloy 600 and stainless steel. 182/82/132 weld metals are used, and these welds are called dissimilar metal welds. These materials have been through cracks due to damage caused by PWSCC (primary water stress corrosion cracking) mechanisms and have caused primary reactor coolant leakage. The damage mechanism was not well reproduced experimentally until the 1970s, and no countermeasures against this damage were taken into account when designing commercial reactors currently in operation around the world.

Therefore, due to this damage mechanism, it is impossible to detect the primary reactor coolant immediately, which raises the question of the effectiveness of the LBB (leak before break) concept used in the recent reactor design. The reason this damage is important is that PWSCC damage of dissimilar metal welds adopted at the primary side boundary of the reactor causes radioactive material to leak out of the reactor.

The PWSCC phenomenon is peculiar to occur in the form of intergranular stress corrosion type (IGSCC) in the primary plant water known as low corrosive water quality. The PWSCC phenomenon is similar to other SCC phenomenon except that the PWSCC phenomenon occurs in the first branched tree, and as the stress increases, the propagation speed increases. In other words, the stress is the driving force of the PWSCC. In fact, the design stress of the nuclear power plant component in which the PWSCC is generated is less than 1/4 of the stress required to reproduce the PWSCC. Nevertheless, since PWSCC is generated, residual stress acts on the site where PWSCC is generated, and at the same time, it is presumed that oxidation or anodic dissolution occurred.

Alloy 600 precipitates a sufficient amount of carbides at grain boundaries after 24 hours at 600 ° C. At this time, the Cr concentration of the grain boundary is lowered, and thus the corrosion resistance of the grain boundary is lowered, which is called sensitization. This explanation is the first explanation for the reason why the grain boundary has low corrosion resistance. Recently, however, it was confirmed that Alloy 600 was the least sensitive in the reactor primary operating environment. This means that the PWSCC phenomenon is not explained solely by grain boundaries. That is, PWSCC of Alloy 600 cannot be explained at least by Cr depletion alone.

In addition, PWSCC is a thermal activation process where the speed of propagation increases as the temperature increases. This means that the PWSCC process involves a reaction that increases in speed with increasing temperature. Experience has shown that this activation energy is reported at 40-50 kcal / mole, depending on how it is determined. However, to date, very few proposals have been validated for the substance of this thermal activation process.

In place of Alloy 600 or Weld 182, which are vulnerable to the PWSCC phenomenon, Alloy 690 and its welding material Weld 152/52 / 52M, which have high resistance to PWSCC, have recently been developed. However, since these alloys are similar to Alloy 600 or Weld 182 in terms of chemical composition, it is reported that a regularization reaction occurs at the reactor operating temperature and PWSCC occurs depending on the conditions.

Thermodynamically, materials mix with each other to increase entropy, which lowers the energy of the system. However, the ordering reaction that occurs only in certain alloys refers to a phenomenon in which a particular atom is thermodynamically stable by repeatedly placing it at a specific position. That is, although specific atoms are uniformly arranged at specific positions due to specific bonds between specific atoms, the entropy is lowered, but the total energy of the alloy system is lowered by increasing the number of bonds between specific atoms.

The regular reactions form a regular array because the regular bonds are stable below the critical temperature (Tc), but when the material is heated above Tc, the regular bonds become unstable and are mixed randomly without distinguishing constituent atoms. It is called. Alloy 600, Weld 182, etc. are considered to have a regularization reaction below 520 ℃ and irregularization reaction at 600 ℃.

It is reported that in the nuclear structural nickel base alloy such as Alloy 600, a regular reaction based on the Ni ₂ Cr (Ni-33 at% Cr) regular structure formed in the Ni-Cr system occurs. In Ni ₂ Cr alloy, when long range order (LRO) is formed by regular reaction, superlattice peak appears and (111) interplanar distance shrinks to 0.4%.

In contrast, the main composition of nickel-based nuclear structure materials, such as Alloy 600, is three elements, Ni, Cr, and Fe, and these alloys form a Ni ₂ (CrFe) structure similar to the Ni ₂ Cr structure. . In this case, the transmission electron microscope showed only fringes, and no superlattice peak appeared. This seems to be because it does not form a long range rule phase such as Ni ₂ Cr, but rather a short range order (SRO). However, the fraction of SRO is high, accounting for more than 80%. This is the reason why the existence of regular reaction is not well known in nickel-based nuclear materials such as Alloy 600.

Therefore, the present invention is to shrink the crystal by suppressing the opening of the grain boundary or stress caused by the shrinkage of the crystal during operation in the primary system environment by allowing the regularization is completed at a temperature lower than 520 ℃ that is higher than the operating temperature of the nuclear power plant and does not occur irregularities However, it is an object to prevent the initiation of PWSCC by suppressing the occurrence of additional stress.

The ordered heat treatment is expected to produce the same effect in the parts or welded parts of the nickel base alloy of the nuclear power plant that was exposed to the primary operating environment of the nuclear power plant without undergoing the ordered heat treatment. Because the rate of normalization at the operating temperature of the nuclear power plant is very slow compared to the regularization process, even when exposed to the nuclear power plant environment prior to the regularization process, the normalization reaction is not completed and there is room for additional regulation. In this case, even if the components are applied to the nuclear reactor primary environment before the regularization process, the regularization will be accelerated and completed.

Unique advantages of the above-described regular heat treatment are as follows. It is known that most materials change the microstructure during processing such as rolling, mechanical processing, welding, defects such as dislocations, and residual stress. If these have a critical impact on the performance of the material or part, they are used after the heat treatment. In addition, many components used in nuclear reactors are used after undergoing heat treatment after processing. However, whether or not the heat treatment is possible can be said that the size of the structure is determined by whether or not the heat treatment is possible (manufacture specification).

In order to solve the problems that occur during processing, the heat treatment temperature after the implementation depends on the materials and components used, but generally needs to be treated at about 600 ~ 700 ℃. However, during heating to this heat treatment temperature, the microstructure of the structure is changed to change the physical properties, or depending on the material causes a fatal problem, such as carbides precipitate at the grain boundary while heating to the post-heat treatment temperature. In addition, nickel base alloys such as stainless steel and Alloy 600 are known to increase carbide susceptibility to corrosion by forming carbides at grain boundaries when maintained at 550 ° C. or higher.

The conditions under which carbides are formed and are sensitive to corrosion are known at 600 ° C for 24 hours for Alloy 600 and depend on the type and condition of the material. Therefore, even if physical heat treatment is possible, heat treatment at 600 ° C. or more is not well selected because it causes significant side effects in addition to the advantages of heat treatment.

In view of this, the ordering treatment at 520 ° C. or below for the representative nuclear power plant structural materials such as stainless steel, low alloy steel, and nickel-base alloys can prevent the initiation of PWSCC while avoiding the microstructure and physical properties of these materials. It is very unique and useful.

The method for preventing PWSCC start of the nickel-base alloy nuclear power plant structural material of the present invention for solving the above problems,

In the method for preventing the PWSCC initiation by regular treatment of the parts including the piping constituting the reactor and the nickel-base alloy which is a welding material connecting the parts, the nickel-base alloy is 0.5 to 200 hours at 400 ~ 520 ℃ To ensure regularization.

The nickel-base alloy is selected from the group consisting of Alloy 600, Alloy 690, Weld 182, Weld 82, Weld 152, Weld 52, Weld 52M, and the ordering reaction rate varies depending on the temperature even in the temperature range of 400 ~ 520 ℃ However, even within the regular temperature range, the heating and cooling is performed at 5 to 10 ° C. per hour in a temperature range of 480 to 520 ° C. where the rate of regularization is relatively high.

In addition, the ordered heat treatment of the parts made of the nickel-based alloy is applied to heat one or more of the inside or outside of the piping parts over the regularized temperature section to maintain the temperature of the other site to the temperature section Oxidizing; The heat treatment is performed in a multi-stage process, including the step of maintaining one side that is maintained at an irregular temperature by applying heat above the temperature section to a regular treatment temperature section in the process of lowering the temperature.

As described in detail above, the method for preventing the first systematic water stress corrosion cracking initiation of the nickel-based alloy nuclear power plant structural material of the present invention completes the regularization of the nickel-based alloy used as a component of a nuclear reactor and exposes it to a nuclear power plant operating environment. By preventing the shrinkage of the additional crystals by regularization in the operating environment of the nuclear power plant, PWSCC is not generated to enable stable operation and use of the nuclear power plant.

In particular, if the inner surface contains a large amount of radiation, such as piping materials used in nuclear power plants, expensive equipment can be heat-treated from the outside without putting it inside the pipe or directly inside, thereby reducing facility maintenance costs and improving work stability. It is possible to provide an improved useful method.

Referring to the method for preventing the first systematic stress corrosion cracking start of the nickel-base alloy nuclear power plant structure of the present invention as follows.

In the method of preventing the PWSCC start by regular treatment of the piping components constituting the reactor and the nickel-base alloy, which is a welding material connecting the components, the nickel-base alloy is maintained by maintaining the nickel-base alloy at 400 to 520 ° C for 0.5 to 200 hours. Let anger be done. The nickel-base alloy is Alloy 600, Alloy 690, Weld 182, Weld 152/52 / 52M. In applying this regularization treatment to the site, a heat treatment of up to 24 hours may be appropriate given the installation or working time of the facility.

The nickel-base alloy is heated at a temperature below 520 ° C. to increase the specific heat at an irregularity, so that the heat treatment is performed at a temperature below 400 ° C .. Most preferably, the ordering rate is increased by shortening the reaction time by heating to a temperature just before the irregularization is achieved.

When the activation energy for the ordering reaction is Q = 46 kcal / mol, the relative rate of normalization according to temperature is calculated according to the rate ∝ exp (-Q / RT) according to the Arrhenius equation and compared with Table 1 below.

Table 1 shows the calculated values and relative ratios of normalization rates from 300 ° C., the nuclear plant operating temperature, to 520 ° C., the regularization temperature. Assuming a nuclear plant operating temperature of 300 ° C, the regularization rate at 520 ° C is about 70,000 times faster than 300 ° C. Above 450 ° C., it can be said that when the ordering treatment temperature rises by about 20 ° C., the ordering treatment speed is about two times faster.

Table 2 shows the time required to complete the ordering according to the relative ordering rate and temperature when the ordering heat treatment time was determined based on 480 ° C using the rate relationship of Table 1. The table was calculated by setting the ordered heat treatment time at 480 ° C. as 0.5, 2, 8, 24, 50 hours. If the regular heat treatment condition of 480 ° C.-1/2 hour is selected, the same effect is obtained when the regular heat treatment is performed for 5.4 to 0.2 hours in the 400 to 520 ° C. interval. In the case of selecting the heat treatment condition of 480 ° C. for 8 hours, the same effect is obtained when the regular heat treatment is performed for 86.72 to 2.91 hours in the 400 to 520 ° C. interval. If the treatment time at 480 ° C is 50 hours, this means that 18.19 hours are required at 520 ° C and 542 hours at 400 ° C. However, low temperature treatment procedures that take a long time to apply to the actual process will not be practical.

In addition, the heating temperature for regular heat treatment is maintained at a constant temperature after direct heating up to the regular temperature, and then cooled. In addition, in the regular temperature range of 400 to 520 ° C., the heating temperature is 5 hours per hour at 480 ° C. The effect will be the same even if the heating and cooling to ~ 10 ℃.

A heat source is required to perform the regular heat treatment, and depending on the shape of the structure of the nuclear power plant, the internal surface must be regularized, but it may be difficult or impossible to reach the heat source in various ways. In this case, the inner surface can be regularly processed by installing and heating the heat source on the outer surface without having to put the heat source on the inner surface. In this case, by using the thermal conductivity of the material, it is possible to calculate the temperature difference between the inner surface and the outer surface, and in consideration of the difference, the inner surface can be heated to a desired temperature.

For example, in the case of the inside of the pipe, since the inside of the primary system pipe is contaminated with radioactivity, there are many difficulties in the work from the inside. In this case, when the pipe is thick or the heat conductivity is bad, the temperature difference between the inner surface and the outer surface may increase. For this reason, in order to set the temperature of the inner surface of the pipe to the temperature range of the present invention, the temperature applied to the outer surface may be defined as a regular temperature range. It may be necessary to heat up to an irregular temperature range. At this time, care must be taken not to increase the temperature on the heat source side to change the fundamental physical properties of the material or to cause sensitization. In order to prevent this phenomenon, when the temperature of the heat source is set slightly lower, the regularization time may be increased.

In this case, the heat treatment is performed first so that regularization is made from the inside of the pipe, but the outside surface of the pipe was applied beyond the regularized temperature range defined in the present invention, so that irregularities were made, and thus the outside surface was maintained at the regularized temperature range. The multi-stage heat treatment may be applied to cool down or cool down gradually so that regularization occurs in the cooling process so that the inner and outer surfaces are regularized.

In other words, the ordered heat treatment of the tubular components made of nickel-base alloy can be heated to the other surface by heating the outer surface or the inner surface. That is, the regularization in the form of piping comprises the steps of maintaining the temperature of the other part in the temperature section by applying heat to at least one of the inside or outside of the pipe in the temperature section; In the process of lowering the temperature, heat treatment may be performed in multiple stages including maintaining one side of the temperature section at which the heat is applied to the temperature section.

Hereinafter, a method for preventing primary systematic stress corrosion cracking initiation of the nickel-base alloy nuclear power plant structural material of the present invention will be described.

DSC analysis of Alloy 600 shows that short range rule is formed. In Alloy 600, SRO is small and does not form a superlattice, so unlike carbide, it is not directly identified. However, when observed by TEM, Alloy 600 appears as fringe, and crystal lattice shrinks according to the progress of regularization, and it is proved by bulk property change or DSC behavior comparison with Ni ₂ Cr regular alloy. The thermal activation energy (Q) for SRO formation in Alloy 600 was found to be Q = 46 kcal / mole.

Nuclear parts manufactured from Alloy 600, Weld 82/182, etc. are subjected to regularization, which is one of thermal activation processes at operating temperature. In this process, lattice shrinkage causes the material to open its own grain boundary or stress during the PWSCC process. Is to generate. That is, since PWSCC is generated by the generation of stress in the material due to the progress of the regularization reaction in the PWSCC process, Q ordering for the regular reaction of Alloy 600 is similar to that of Q pwscc.

Figure 1 shows the specific heat change of Alloy 600 by DSC analysis, and shows the results of the thermal analysis performed after various different thermal and mechanical treatments for Alloy 600. At 600 ℃ or higher, water quenching (WQ) materials generate exothermic reactions due to regularization at around 450 ℃, whereas cold work (CW) materials can be subjected to regularization even between about 100 ~ 300 ℃. An exothermic reaction occurs. This means that cold working makes alloy 600 more sensitive to PWSCC by lowering the ordering temperature of Alloy 600 and providing additional drive for regularization. Although the ordering temperature varies depending on the treatment conditions of the material, it can be seen that the ordering reaction occurs at 520 ° C. or lower irrespective of the treatment conditions, whereas at 520 ° C. or higher, irregularity occurs as the specific heat increases.

2 is a graph showing the change in distance between the crystal plane after the isothermal annealing treatment at 400 ℃. Where a) is the change of the interplanar spacing determined by neutron diffraction analysis after isothermal annealing treatment of the WQ test specimen at 1100 ° C. at 400 ° C., and b) is a 25% cold rolled AC test piece, followed by isothermal annealing treatment at 400 ° C. It is a change in distance. Here, it can be seen that the shrinkage of the crystals is anisotropically different depending on the direction by the regularization treatment at a temperature higher than the nuclear operation temperature.

Shrinkage due to regularization depends on the treatment conditions and the crystallization direction and the size is about 0.02 ~ 0.07% in WQ material above 600 ℃. This degree of shrinkage has the effect of spreading the grain boundary up to several tens A (angstrom) in consideration of the grain size, and the grain boundary having a large difference in shrinkage ratio due to the large crystal orientation difference will tend to open more. This can also be explained from the viewpoint of stress, and additional stress due to crystal shrinkage can be obtained from the shrinkage ratio and the modulus of elasticity according to Young's Law, and is about 40 to 150 MPa depending on the combination of grain boundaries.

The angle between the [111] and [311] directions is 29.5 2 and the difference in shrinkage between these two directions in the water quenching material is the largest, about 0.07%, and reaches 150 MPa in terms of stress. The angle between the [111] and [220] directions is 35.3 3, and the difference in shrinkage between these two directions is about 0.06%, which is about 130 MPa in terms of stress. In addition, the angle between the [311] direction and the [200] direction is 25 mm, and the difference in shrinkage according to the regularization is about 0.05% at maximum, which translates into about 110 MPa.

In addition, as a result of observing the effect of ordering treatment on Alloy 600, which is not shown, but treated with various treatments such as WQ, AC, FC, CW, etc., Alloy 600 has different shrinkage behavior of crystal depending on pretreatment conditions. It can be seen that it contracts during the regularization process.

Tensile testing is a useful method for testing the effects of deformation in real time while inducing deformation at the desired deformation temperature. 3 is a graph showing the change of the distance between crystal lattice planes according to the tensile strain temperature after neutron diffraction analysis of the tensile strain test specimen at a strain rate of about 10 ^-3 / sec at various temperatures from room temperature to 750 ℃. As shown in FIG. 3, the neutron diffraction results of the tensile strain test specimens showed that the distance between the crystal planes was reduced at about 150 to 600 ° C. similarly to the DSC results. The DSC results shown in FIG. 1 show that the regularization occurs at 50 to 520 ° C., while at 150 to 600 ° C. in FIG. 3, the strain strain rate is as fast as 10 ⁻³ / sec and the tensile strain continuously during the ordering reaction. This seems to be due to the difference being applied. That is, the regularization phenomenon shown in FIGS. 1 and 3 is essentially the same.

This means that deformation at 500 ° C. or lower promotes regularization according to stress, causing shrinkage. Therefore, it can be explained that the deformation of the crack tip causes regularization during the PWSCC process, and that the contraction action controls the crack propagation process. The crystals in the [111] and [200] directions shown in FIG. 3 are 54.7 mm, the angles in the [220] and [200] directions are 45 mW, and the angles in the [111] and [220] directions are 35.3 mm, respectively. It shows that the shrinkage difference between them is the largest. This is in good agreement with the experimental observation that PWSCC occurs at high-angle grain boundaries.

As described above, the ordering process includes thermal ordering, which occurs in a pure thermal activation process without stress or deformation, and strain ordering, in which ordering is rapidly advanced by the action of deformation. The thermal regularization described in FIG. 1 is important at the crack generation stage of the PWSCC, and the deformation regularization described in FIG. 3 is important at the PWSCC crack propagation stage.

The Alloy 600 will form a Ni ₂ (CrFe) SRO of Ni ₂ Cr type in a reactor operating temperatures of around 300 ℃. This regularization reaction rate increases exponentially with increasing temperature and becomes faster when stress is applied. During the formation of SRO, the shrinkage of the lattice occurs because the distance between atoms between specific atoms is shortened, and the amount of shrinkage is a measure of ordered progress.

4 is a schematic diagram showing a process in which local stress acts on the grain boundary due to anisotropic shrinkage of the crystal due to SRO formation, a) is an initial state in which processing is completed and no regularization occurs, and b) is according to regularization The effect of shrinkage of the crystal is shown schematically.

In the above process, crystals having similar grain orientations have similar contraction directions to cancel the contracting effect, and grain boundaries having a large difference in shrinkage ratio due to large orientation differences of the crystals are opened according to regularization. When shrinkage occurs in directions opposite to each other due to the combination of the orientations of adjacent crystals, the opening of the grain boundary or the magnitude of the stress can be amplified by about twice. In this case, the magnitude of the stress generated at the high-angle grain boundary can be increased up to 300 MPa, and the gap of grain boundary can be expanded up to 40A when the grain size is about 50 μm, and can be expanded up to 100A by the action of the surrounding crystal. Can be. Since the size of water molecules is about 5 A, nuclear power medium can penetrate between grain boundaries. Therefore, PWSCC sensitivity is highly dependent on the presence of grain boundary phenomena due to shrinkage of crystals that occur while being ordered in the state exposed to the first order of high temperature and high pressure.

In addition, the grain boundary must include a triple point where several crystals meet, and the cross section of this point is a point, but this part is mostly high-angle grain boundary because at least three crystals with different crystal orientations meet. . These vertices are spatially connected along the interior of the crystal, which is the most vulnerable when the crystal contracts. Thus, if the interstitial grain spacing of the crystal due to the contraction of grain boundaries or spatially connected vertices is wider than the water molecules, the primary cooling water will not only contact the surface of the material but also penetrate into the interior of the material.

In the PWSCC cracking stage, when the alloy 600 is exposed to the reactor environment, the anisotropic contraction causes the compatibility to be broken first, and thus the pressure effect of the first nuclear power plant is applied first. At this time, since the difference in shrinkage is large at grain boundaries where the orientation difference is large, the high-angle grain boundary is partially opened first, and the portions are slowly connected to develop cracks. If it stops without growing because of a large defect, it is a damage called intergranular attack (IGA).

The shrinking process according to the regularization acts not only at the stage of crack development but also at the propagation stage. The defect formed as described above may be judged to have entered the crack propagation stage when a plastic station is formed at the crack tip due to the stress concentration effect caused by the movable stress. At this time, the formation of the plastic zone means that the local plastic deformation of the crack tip is formed, and since the plastic deformation promotes the formation of SRO, the stress generation or grain boundary phenomena caused by the formation of SRO acts once more (see FIG. 3). ). In other words, when the firing station starts to be formed once at the tip of the crack due to the movable stress, the material itself forms an SRO in the firing station so that the grain boundary is opened, so that the crack growth continues and grows as a through crack. Shrinkage by SRO formation in the firing zone at the crack tip is controlled by the regular reaction rate, which appears to depend on stress. That is, it depends on the interaction of temperature and stress. Due to this relationship, Q pwscc propagation = 40 to 50 kcal / mol is similar to Q ordering = 46 kcal / mol.

Conventionally, PWSCC of Alloy 600 considers only the residual stresses in explaining the corrosion phenomenon and does not consider the effects of the stresses generated in the PWSCC process. The difference is that nickel-based alloys change through a regularization process, and that the change in this material causes grain boundaries or additional stresses to be understood and controlled. That is, PWSCC is a phenomenon that is controlled by the generation of additional stress or grain boundary opening due to SRO formation according to regular reaction. This is the substance of the new PWSCC mechanism, in which the stress source operates and propagates.

PWSCC is considered to be controlled by intergranular corrosion because IGSCC type and primary plant water environment is essential. However, there is a limit in that the PWSCC phenomenon cannot be adequately explained by the oxidation of grain boundaries alone. For example, it was not possible to explain the occurrence of PWSCC even if the working stress was much lower than the stress required to reproduce the PWSCC. In order to overcome this limitation, the existence of residual stress has been assumed, but the residual stress is understood to have a fixed value once it is formed with the maximum value at the beginning of nuclear power plant operation or to decrease gradually with the operation of nuclear power plant. Inappropriate to explain the activation process. Furthermore, there may be limitations that may exist in the PWSCC process but cannot be verified.

However, the basis of the present invention is that the PWSCC phenomenon satisfies the sufficient conditions because the PWSCC itself supplies the necessary stresses because the PWSCC phenomenon causes additional stress as the material changes as it is thermally activated, or causes the grain boundary to open while shrinking by itself. In other words, this description can fundamentally explain the thermal activation process appearing in the PWSCC, so it is expected that at least the PWSCC phenomenon will replace the existing grain boundary carbide and grain Cr depletion.

FIG. 5 compares concepts representing a combination of stress, environment, and material conditions in which PWSCC occurs. Figure 5a) is a conceptual diagram of the existing model, Figure 5b) is a conceptual diagram of the model proposed in the present invention. In FIG. 5 a), the existing model shows that PWSCC occurs when the composite effect of stress, environment, and material is specified, whereas the basic model of the present invention shown in FIG. The PWSCC is generated only when the specific condition of generating a is satisfied. Qualitatively, PWSCC occurs in nickel-based alloy parts, and high-angle grain boundaries exposed to the nuclear power plant environment are essential.

According to the model, Alloy 600 or Weld 182 material is employed as the reactor component, and regularization occurs during operation, creating additional stresses, resulting in PWSCC. However, until now, the PWSCC could not be controlled because of insufficient understanding of the PWSCC mechanism and the control of Alloy 600 or Weld 182, and thus the risk of the PWSCC. However, according to the underlying model of the present invention, since the generation or propagation of PWSCC is controlled by the shrinkage of crystals generated during the regularization process, the regularization of Alloy 600 or Weld 182, which is used as a reactor part, is completed. Exposure to a water environment prevents the initiation of PWSCC.

Considering the effect of the stress formed in the regularization process and explaining the process of increasing the stress as follows. Except for the design stress, the combination of the presence and absence of various stresses is so extensive that only five types of stresses are considered here. These stresses are a. Operating stress due to pressure in the reactor, b. Stress due to internal and external temperature differences of nuclear power plant components, c. Residual stress, d. Additional stresses arising from the regularization process resulting from thermal activation during the cracking stage, e. It is an additional stress caused by the strain induced ordering process in the propagation phase of the crack.

Here, the stress d and the stress e do not have a constant value but vary depending on the orientation difference of the grain boundaries and thus have a certain range. This is because it depends on the crystal orientation difference of the grain boundaries exposed to the reactor primary cooling water as described above. If the shrinkage difference between the grain boundaries is not large, this value is low, which is why the PWSCC is only observed at the high grain boundary because the low grain boundary is not vulnerable to the PWSCC. The meaning of the high-angle grain boundary narrowly means not only the angle between two grains, but also the region where these high-angle grain boundaries are densely packed. Only in this case, unlike IGA, it is expected to grow into PWSCC.

When the reactor is running, at least stresses a and b act, which is less than one-quarter of the critical stress needed to produce the PWSCC. However, if the residual stress (stress c) is sufficiently high, it is expected that PWSCC may occur since the sum of these stresses exceeds the critical stress of the PWSCC. In this case, it is expected to occur early in the operation of the nuclear power plant. On the other hand, the residual stress decreases with time, but it may be possible to interpret that through cracks are formed decades after penetration cracks are formed by oxidation or corrosion.

However, considering that PWSCC is a passion activation process, the above oxidation or corrosion process should be able to explain Q pwscc = 40 ~ 50 kcal / mol. Moreover, even if residual stresses generate PWSCC, the ordering process that proceeds slowly in nickel base alloys during the operation of the reactor is inevitable. Therefore, even if the residual stress is low or absent, the PWSCC is initiated when the additional stress, d, at a particular grain boundary of the nickel base alloy in the primary water is high enough to reach the PWSCC critical stress. For this reason, Q pwscc initiation = 40 to 50 kcal / mol is similar to Q ordering = 46 kcal / mol.

If the defect does not grow to a sufficient length at this stage, it is only IGA damage, and if it can grow further under the action of moving stress, residual stress, and regularization stress, it enters the crack propagation stage and a plastic zone is formed at the crack tip. do. In this plastic zone, deformation occurs and dislocation propagation occurs, so that the plastic zone undergoes strain organic regularization as shown in FIG. 3. This means that in the propagation phase the grain boundary opening is controlled by the regularization rate, so that the PWSCC crack propagation is controlled by the regularization process. Due to this relationship, Q pwscc propagation = 40 ~ 50 kcal / mol has similar value to Q ordering = 46 kcal / mol.

As described above, in the past, the existence of the regularization process was not recognized, and thus the effect of the rule reaction occurring in the nuclear power plant operating environment could not be considered. The basis of the present invention, however, is that the nickel base alloy causes a regularization reaction in nuclear power plant operating conditions, whereby the grain boundary gap or additional stress is controlled by generation.

The existing model for describing a PWSCC is that a PWSCC occurs if only three elements are met: environment, material, and stress (moving and residual stress). This model is inadequate to explain why PWSCC occurs at high-angle grain boundaries, why PWSCC occurs at relatively low stresses, and thermal activation processes. On the contrary, the basic model of the present invention differs from the existing model in that it requires additional requirements of the regularization process and the region of the high angle boundary.

The basic model of the present invention requires the environment, materials, stresses (moving and residual stresses), regularization (grain opening or additional stress), and exposure conditions of dense regions of high-angle grain boundaries. Among these, only the regularization process can be controlled by the regularization process. If the role of the regularization process is important in the PWSCC process, then the PWSCC may be controlled if the control process is controlled.

According to the basic model of FIG. 5 b) of the present invention, it can be expected that the PWSCC does not occur unless regularization occurs in the primary water environment of the nuclear power plant. Thus, a comparison of PWSCC initiation with a sensitive and ordered nickel base alloy material that is sensitive to PWSCC may reveal whether the ordering treatment prevents PWSCC initiation. This is because, as described above, the only adjustable PWSCC requirements are the regularization process control.

As a result of measuring residual stress on the ordered heat treatment test specimen using Cu Ka X-ray, the residual stress was reduced by the ordering treatment. However, as shown in FIG. 2, since the crystal lattice by the regularization treatment shrinks, it is not possible to clearly distinguish whether the reduction of the residual stress due to the regularization treatment is an effect of the crystal lattice shrinkage or the reduction of the residual stress due to the regularization treatment. Can't. This is because the residual stress measured by the XRD method uses the principle of measuring the residual stress by measuring the interplanar distance of the crystal. In other words, it can be said that the effect of preventing PWSCC initiation caused by the regularization treatment is that the reduction of residual stress and the regularization effect acted simultaneously. However, since the normalized treatment temperature of 480 ° C is considered to be relatively low for residual stress reduction due to the annealing effect, such a change may be due to the existence of a regular reaction even when the residual stress decreases due to the regularization treatment. can do.

Table 3 shows the residual stress reduction effect of the CW Alloy 600 heat treated at various temperatures for 2 hours and measured in the loading direction of the SSRT specimen (FIG. 6). In the table,-means decrease and + means increase. In Table 3, the residual stress was found to decrease only before the heat treatment, but only 480 ° C-2 hours. The residual stress did not decrease or increased due to the heat treatment above 550 ° C. Table 4 shows the effect of reducing residual stress at the ordered treatment temperature of 480 ° C., which shows that the residual stress decreases at least as before or as a whole. The change in reduction shown in Table 4 is reasonable considering the characteristics of residual stress measurements.

The residual stress reduction measured in the load direction of the SSRT test piece (Fig. 8) after heat treatment at various temperatures above the ordered treatment temperature for the Weld 182 is shown in Table 5. As can be seen from Table 5, the treatment conditions in which the residual stress is reduced than before treatment are only 480 ° C-2 hours. The residual stress was found to decrease by up to 1500 MPa. Even if the heat treatment temperature was increased above 550 ℃, the residual stress was slightly decreased or increased.

Table 6 shows the reduction of residual stress with regular heat treatment time. Although there is a slight deviation with the increase in the ordered heat treatment time, it can be seen that about 350 MPa decreases even when only 1/2 hour treatment is performed at 480 ° C., and a reduction effect of 1500 MPa or more occurs as the time increases.

The reduction of residual stress in cold rolling direction of CW Alloy 690 according to heat treatment conditions is shown in Table 7. After heat treatment at 480 ℃ for 2 hours, residual stress decreases by about 280 MPa. However, residual heat decreases by about 26 to 180 MPa when heated at 550 ℃ for 2 hours.

As shown in Table 8, the regular heat treatment reduces the residual stress under all treatment conditions, and the effect of reducing residual stress from 125 MPa to 278 MPa is obtained over time.

As shown in Tables 3 to 8, it can be seen that the effect of reducing residual stress is mainly caused by regular heat treatment, and the heat treatment is performed without causing a change in microstructure and physical properties due to an increase in temperature. It is very useful from the viewpoint of suppressing deformation induced by the structure due to thermal expansion of heavy materials or preventing the material from sensitization by precipitation of carbides.

Therefore, the effects of the above invention can be achieved through the low strain rate test (SSRT) conducted in the first-order environment of nuclear power plants before and after the regularization treatment for Alloy 600 and Weld 182 in the cold working state. Verification was possible, and the regularization process prevented the initiation of PWSCC. The verification process is described in detail below.

Experimental Example 1

It can be seen that Alloy 600 occurs mainly at the expansion site in the case of steam generator tubing and Weld 182 is itself sensitive to PWSCC. Therefore, the steam generator tubing can be rolled at room temperature, making it vulnerable to PWSCC, and then the effect of regularization can be demonstrated by comparison with the regularized specimen. In addition, when the Weld 182 is fabricated to make a specimen, and the PWSCC comparison experiment with the regularized specimen can be proved that the regularization treatment can prevent the PWSCC initiation.

In this experiment, whether the primary water quality is adequately satisfied is an important factor. Therefore, in this experiment, it was verified that the primary water quality is properly maintained by monitoring the primary water quality that is returned around the specimens by creating a circulation loop that can maintain the same water quality as the nuclear water environment.

Summarizing the water environment, pressure is 160 bar, temperature is 330 ° C, boric acid concentration is 1200 ppm B, Li concentration is 2.2 ppm Li, dissolved oxygen (DO) is 10 ppb or less, dissolved hydrogen (dissolved hydrogen) ) Is around 2 ppm and the conductivity is about 20 μS / cm.

The experimental method used in the present invention used a method of comparing the crack initiation characteristics of the surface using a method of slow deformation in the primary water environment. This test method is known as SSRT, and a photograph comparing the surface cracks of CW Alloy 600 and regularized Alloy 600 specimens with strain rates of 3 x 10 ^-8 / s ^{-1 up} to 2.7% is shown in FIG. It was.

After the SSRT test, the entire area of the gage section was examined by a scanning electron microscope (SEM). As shown in FIG. 6A, the material rolled at room temperature showed grain boundary cracking by the test. Cracks were not found in the 480 ℃-8 hour ordered material shown in b).

Therefore, in the cold processed Alloy 600 material, not only cracks were observed on the surface but also the surface was more vulnerable than the regularized specimens. Therefore, it can be seen that the regularization treatment prevents the initiation of PWSCC. Can be.

Experimental Example 2

In order to verify the effect of regularization on Weld 182, which is known to be sensitive to PWSCC, Weld 182 test piece having a cross section as shown in FIG. 7 was fabricated. After the normalization treatment on the Weld 182, the surface crack photographs of the Weld 182 and the regularized Weld 182 specimens subjected to SSRT test at 2.7% with a strain rate of 3 × 10 ⁻⁸ / s ⁻¹ were compared with those of FIG. 8.

As shown, the Weld 182 (FIG. 8 a) that was not ordered showed grain boundary cracks along the dendrite cell boundary, whereas the weld 182 (FIG. 8 b) which had been treated with 480 ° C.-2 hrs was cracked. Does not occur. Although the photograph is not presented, it can be seen that there is a crack initiation prevention effect even when the 1/2 hour treatment at 480 ℃, and the regularization reaction is expected to be completed even at 480 ℃-0.5 hour.

The regularization treatment mentioned here is characterized in that the treatment method is the same as a general heat treatment method, but the purpose or treatment temperature section is a temperature section in which regularization progresses. Although described with reference to FIG. 1, irregularity occurs at 520 ° C. or higher, and the crystal lattice expands in this process. In this case, when the nuclear power plant parts are operated at the operating temperature of the nuclear power plant, regularization occurs again, causing the shrinkage of the crystal lattice, and thus, the PWSCC is sensitive again.

On the other hand, in the case of heat treatment at 520 ℃ or less, the regularization progresses and the shrinkage of the crystal lattice is completed. Therefore, when the reactor is exposed to the primary environment of the reactor, no further regularization process occurs, and thus, PWSCC cannot provide sufficient conditions. Since the initiation is prevented, the regularization treatment should be distinguished from the general heat treatment.

As shown in FIG. 1, the regularization appears to occur at 520 ° C. or lower, but can be said to occur at 500 ° C. or lower, and the PWSCC prevention is prevented when the regularization treatment time is processed at 480 to 500 ° C. for 30 minutes or more. It seems to be.

However, in order to apply to actual nuclear power plant components, at least several hours of processing time will be required in consideration of the time required for heating and heat transfer. In this way, when the heat source is installed on one side and the other side is subjected to regular processing, if the thickness of the part is thick and the thermal conductivity is low, the temperature of the outer surface must be selected slightly higher. After the regularization process is completed, the exterior surface can be processed in two stages. The temperature-time relationship in the multi-step heat treatment to this process is schematically shown in FIG.

The present invention has been developed to alleviate the PWSCC of the nickel-base welded to date, and the laser peening, water jet peening, and weld inlay methods listed in the ASME code case require direct treatment on the surface of the nuclear power plant. Therefore, these methods suffer from practical problems such as access limitations due to the complex structure of the nuclear power plant, exposure to radiation, precise treatment of the mitigation portion, and monitoring. It is possible to neglect the internal mitigation without having to deal with the parts.

In addition, the above-mentioned example is only an example to demonstrate this invention. Therefore, it is obvious that the ordinary skilled in the art to which the present invention pertains uses the partial change with reference to this detailed description.

1 is a graph showing the specific heat change of Alloy 600 by DSC analysis.

Figure 2 is a graph showing the change in distance between the crystal plane after isothermal annealing treatment at 400 ℃ the test piece.

Figure 3 is a graph showing the distance between the crystal lattice plane to the strain temperature by neutron diffraction analysis for the tensile strain test specimen at various temperatures.

4 is a schematic diagram illustrating a process in which local stress acts on grain boundaries due to anisotropic shrinkage of a crystal due to SRO formation;

FIG. 5 is a conceptual diagram comparing concepts representing a combination of stress, environment, and material conditions in which PWSCC is generated. FIG.

Figure 6 is a comparison of the surface cracks of SSRT-tested cold processed Alloy 600 and regularized Alloy 600 test piece.

7 is a schematic view of Weld 182 welded specimen preparation.

Figure 8 is a comparison of the surface cracks of the SSRT test Weld 182 and regularized Weld 182 test piece.

9 is a temperature distribution diagram of a multi-stage heat treatment process required when the side to be normalized and the side to be in contact with the heat source are different.

Claims (4)

In the method of preventing the first systematic stress corrosion cracking initiation of the nickel-base alloy nuclear power plant structural element to prevent the PWSCC start by the regular treatment of the nickel-base alloy piping parts constituting the reactor and the nickel-base alloy that is a welding material connecting the parts , A method for preventing primary strain stress corrosion cracking initiation of a nickel-base alloy nuclear power plant structure, wherein the nickel-base alloy is maintained at 400 to 520 ° C. for 0.5 to 200 hours. The method of claim 1, The nickel-based alloy is selected from the group consisting of Alloy 600, Alloy 690, Weld 182, Weld 82, Weld 152, Weld 52, and Weld 52M. Prevention method. The method of claim 1, Method for preventing the first system strain stress corrosion cracking start of the nickel-base alloy nuclear power plant structural material, characterized in that the heating and cooling is performed at 5 ~ 10 ℃ per hour in the temperature range of 480 ~ 520 ℃ well represented in the temperature section. The method of claim 1, The regularization of the pipe is a part made of the nickel-base alloy Applying heat to at least one of the inside and outside of the pipe to the temperature section to maintain the temperature at the other part at a temperature range of 480 to 520 ° C .; Primary systematic stress stress corrosion of the nickel-base alloy nuclear power plant structural material comprising a multi-stage heat treatment process including the step of maintaining one side of the heat applied to the temperature section in the process of lowering the temperature in the temperature section How to prevent crack initiation.

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