JPH0649915B2 - Stainless steel for nuclear reactor core equipment and manufacturing method thereof - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION [Industrial field of application] The present invention relates to the amount of cobalt 58 produced by the activation of nickel eluted from a material in a reactor in a reactor in contact with high temperature water. The present invention relates to a reactor core material material in which nickel is suppressed to the minimum necessary in terms of alloy composition in order to reduce the amount and a manufacturing method thereof.

[Conventional technology]

Various austenitic stainless steels having excellent corrosion resistance are used for conventional nuclear reactor core material. Generally, austenitic stainless steel is often used for nuclear reactors.

In particular, in the case of nuclear reactor core material, attention must be paid to C content (%), Ni (%), Cr (%), and N (%) as a trace element.

Among these, nickel is a stabilizing element of austenite, and as a basic element of austenitic stainless steel, for example, in JIS standard G4304 SUS304 steel, approximately 6 wt% to 1 wt.
It is customary to contain about 0.5 wt%. In other austenitic stainless steels, nickel is
Some even include wt% or 22 wt%. Further, some nickel-based systems, such as NCF600, contain 72 wt% of Ni.

In equipment composed of these alloy materials, the parts that come into contact with the reactor coolant such as high-temperature water are corroded by the cooling, and some of the alloy composition elements are dissolved in the coolant. May be accumulated as corrosion products.

When these corrosion products pass through the reactor core, they are irradiated with thermal neutrons, and the cobalt and nickel contained in the corrosion products are transmuted into radioactive cobalt 60 and cobalt 58, respectively. . Since these radioactive elements emit gamma rays deposited in various parts of the reactor core and pipes over a long period of time, it is necessary to take radiation protection measures for workers in maintenance and inspection and repair work of reactor equipment. The efficiency may decrease.

In order to deal with this, the cobalt content in the austenitic stainless steel is usually reduced from the normal content of 0.2 to 0.4 wt% mainly for the purpose of reducing cobalt 60 having a long half-life.
Measures have been taken such as limiting the content to 0.01 to 0.05 wt% or less. (For example, Japanese Patent Application No. 51-26756). As a result, the amount of cobalt 60 generated in the present nuclear reactor has been greatly reduced.

However, nickel has been difficult to reduce because it is a basic element of austenitic stainless steel. Therefore, as a measure to reduce the amount of cobalt 58, a small amount of iron ions and iron oxides are injected into the reactor feed water, and the cobalt 58 produced in the reactor is taken in as iron oxides and stably adhered to the surface of the fuel rod. Measures such as allowing them to be taken are being taken. However, these methods cannot be a method of fundamentally reducing the amount of cobalt 58, and therefore, there has been a drawback that radioactivity treatment measures are required in steps such as fuel reprocessing.

On the other hand, as a method of drastically reducing cobalt 58,
In place of austenitic stainless steel, there is a movement to construct nickel core-free stainless steel for reactor core equipment by using ferrite stainless steel. However, although ferrite materials are used for pressure vessels, they are not generally used from the viewpoint of irradiation embrittlement resistance for the reactor core equipment with large thermal neutron irradiation dose. It was

Further, a method of improving the irradiation resistance by containing N more than impurities is being studied, but N is 0.0
If it is 8 wt% or more, the weldability is deteriorated, so that there is a drawback that it is difficult to apply to equipment having a welded portion.

[Problems to be Solved by the Invention]

The above-mentioned prior art does not have an effective measure for reducing the amount of cobalt 58 in the austenitic stainless steel because the Ni content is large, and in the case of the ferrite stainless steel, there is a problem in workability such as weldability. It was

An object of the present invention is to suppress the elution of nickel in the coolant by using a low nickel type stainless steel with the minimum nickel content suppressed as the reactor core material, and to suppress cobalt 58 by activation. An object of the present invention is to provide a reactor core material and a manufacturing method thereof, in which the generation of heat is suppressed.

[Means for Solving the Problems]

The structure of the stainless steel for reactor core equipment according to the present invention for solving the above-mentioned problems is a material used for manufacturing a reactor core equipment,
Fe as a main component, C, Si, Mn, P, S, Ni, C
In an austenitic stainless steel containing r, Mo, Cu, Al and impurities, Ni = 1 to 6%, C <
0.03%, Cr = 12 to 20%, N is C + N <
It was added in an amount of 0.05% so as to pass the U-bend corrosion test with a gap in high temperature and high pressure water for 1000 hours.

Further, the structure of the manufacturing method is such that Fe, which is used as a main component for manufacturing reactor core equipment, is composed of C, Si, Mn,
In the method for producing austenitic stainless steel containing P, S, Ni, Cr, Mo, Cu, Al and impurities, Ni = 1 to 6%, C <0.03%, Cr = 12 to.
20%, N is added so that C + N <0.05%, melt casting is performed, and this cast product is subjected to 1 at 950 ± 10 ° C.
~ After heating for 3 hours and quenching, 600 ~ 10 ℃ 3 ~
The tempering heat treatment is performed for 10 hours.

[Action]

As shown in FIG. 7, the base material of the test material obtained by subjecting the thick plate to V-shape welding and the weld metal, respectively, to the bending test piece with a gap shown in FIG. 8 and the tensile test piece shown in FIG. Were collected and subjected to a stress corrosion cracking test. The stress corrosion test was carried out by dipping in high temperature pure water at 320 ° C. and 288 ° C. for up to 1000 hours. In the U-bend test with a gap, stress corrosion cracking usually occurs only on the back side (tensile stress side) of a U-shaped bending test piece. Further, in the low strain rate tensile test, if it is carried out while being immersed in high temperature water, the presence or absence of the intergranular stress corrosion cracking susceptibility can be known by observing the fracture surface after fracture.

The U-bend test with a gap is carried out for the purpose of detecting stress corrosion cracking of the welded member and can be said to be the most severe judgment method. On the other hand, the low strain rate tensile test is a kind of tensile test, and is a convenient method for observing the fracture surface of the test piece after the test. In the present invention, the former is adopted for the determination of stress corrosion cracking, which is the most important issue for austenite. In the following explanation, the conclusion was reached by observing the presence or absence of cracks or the fracture surface ratio of these test pieces.

Hereinafter, these results will be described in detail.

C is an element that remarkably enhances the susceptibility to stress corrosion cracking of stainless steel in a high temperature pure water environment. In the case of stainless steel, C content of 0.03% or less can significantly reduce the stress corrosion cracking susceptibility. The technique for reducing the C content has been remarkably improved due to the recent progress in the technique for refining in the ladle, and the above can be achieved.

Si is an essential element as a deoxidizer during steel making,
There is almost no effect on the improvement of stress corrosion cracking in a high temperature pure water environment. Rather, if it exceeds 0.75%, the formation of intermetallic compounds such as σ phase is accelerated, resulting in deterioration of workability and weldability. Therefore, the upper limit is set to 0.75%.

Mn is also an indispensable element as a deoxidizing agent during steel making, but it hardly affects the stress corrosion cracking resistance in a high temperature pure water environment.
Further, Mn is an austenite forming element, and if it exceeds about 1%, the martensite structure and the ferrite set become unstable, so the upper limit is about 1%.

Since P and S contained as other impurities adversely affect weldability, workability, etc., P about 0.030% and S about 0.
It is necessary to suppress it to 030% or less. (Further, in order to improve the corrosion resistance in the vicinity of grain boundaries due to the segregation of impurities due to neutron irradiation, P is 0.010 wt% or less and S is
Is preferably 0.005 wt% or less).

Cr is an indispensable alloying component as a corrosion resistance improving element in stainless steel, and at least 12% or more is necessary to maintain the corrosion resistance in a high temperature water environment. In high temperature pure water such as reactor water, there is almost no salt containing Cl ^{− and the} like, so Cr is at least 12%.
If so, there is no problem in corrosion resistance. Further, as the amount of Cr increases, the corrosion resistance improves, but on the other hand, when the amount of Cr increases, the workability deteriorates, leading to an increase in cost, so the upper limit is approximately 20.
%.

Mo is an extremely effective element for reducing the local corrosion generated in the gap portion formed when stainless steel is used, that is, the crevice corrosion. Even a small amount of 0.1 wt% is effective for repassivation of the surface film, and the addition of 2.5% can improve the low temperature toughness. However, if it exceeds 4.0%, an intermetallic compound such as a chi phase tends to be generated and workability and weldability are significantly deteriorated. Is 2.5%.

Cu is a component necessary for improving acid resistance, and its effect is shown by 0.05% or more. Further, they contribute as Cu interstitials during neutron irradiation, and have an effect of suppressing impurity segregation and Cr depletion. But 1
%, Cu intermetallic compounds precipitate, which rather deteriorates acid resistance. Therefore, the upper limit is 1%.

Al is necessary for improving the toughness of the welded portion, particularly by deoxidizing the steel and refining the crystal grains. At least 0.05
% Is not contained, there is no effect of refining the crystal grains. Further, from about 0.2%, an effect of preventing cold cracking at the time of welding also occurs. However, if it exceeds 1%, the creep property is deteriorated, so the upper limit is made 1 wt%.

In particular, reduction of the amount of Ni added is one of the aims of the present invention. Regarding the lower limit of the amount of Ni, Ni is effective in improving the low temperature toughness, and the amount is required to be 1.0% or more.

In addition, 1.0% or more addition is required to stably and densely form a surface oxide film of stainless steel in high temperature water. On the other hand, the upper limit of the amount of Ni added is preferably as small as possible. If the amount of Ni is increased, the crystal structure is changed to an austenite phase, and improvement in strength due to heat treatment cannot be expected. It was set to 6.0%.

Although N is effective in increasing the high temperature strength, it has been found that ferrite and martensitic stainless steels reduce the stress corrosion cracking resistance.
FIG. 6 is a relationship diagram between (C + N)% and weld crack ratio (%) of 12Cr steel. (C + N)% is 0.05 wt%
It is clear that the weldability deteriorates sharply in the above cases. Therefore, the upper limit of C + N is 0.05%.

Next, the heat treatment conditions for the steel of the present invention will be described.

(A) Regarding Quenching Conditions FIG. 4 is a characteristic curve diagram of the quenching temperature and mechanical properties (hardness and absorbed energy) of the invention steel A.

As can be seen from FIG. 4, the quenching temperature is 900 ° C to 100 ° C.
The hardness (H _B ) is stabilized at 0 ° C. at 250 to 260, and the absorbed energy is also stabilized at a value of 16 to 18 (kgf · m). Here, 950 ± 10 ° C. was selected as a suitable quenching temperature in consideration of ensuring uniform heat treatment at the work site. When the holding time is 50 minutes or more, the hardness decreases rapidly,
From the result (not shown) that the hardness increases again from time, 1
~ 3 hours was chosen as the preferred quench hold time.

(B) About tempering conditions FIG. 5 is a characteristic curve diagram of tempering temperature and mechanical properties (hardness and absorbed energy) of the steel A of the present invention.

In FIG. 5, the tempering temperature decreases in hardness from 580 ° C to 590 ° C, and the absorbed energy increases. Also, 6
At 10 ° C or higher, hardness increases and absorbed energy decreases. From the above results, as a suitable tempering temperature, 610
Suggest ± 10 ° C. Furthermore, when the tempering holding time is 3 hours or more, the hardness is lowered and the absorbed energy is increased (not shown). From 8 hours to 20 hours, almost constant hardness and absorbed energy value can be obtained, so considering the reduction of heat treatment cost at the site, the tempering holding time is 3 to 1
You can propose 0 hours.

〔Example〕

Examples of the present invention will be described below with reference to Tables 1 to 3 and FIG.

Table 1 is a list of the chemical composition of the test material and the stress corrosion cracking resistance (SCC) resistance in the U-vent test with a gap in the welded portion.

In Table 1, the comparative steels (G, H, I, J) have the chemical composition in the material of the conventional example, while the invention steels (A,
B, C, D, E, and F) have the chemical composition C, Si, ... It is limited to the range of 6.0% and the amount of (C + N) is limited to 0.05% or less. In the following examples, invention steels (A, B, C,
D, E, F).

As a result, in the invention steel, (1) crystal grains became finer and (2)
Good weldability was exhibited, and (3) the welded test piece in high-temperature high-pressure water did not cause bending cracks and exhibited good SCC resistance. On the contrary, in the comparative steel of the conventional example, cracks were observed in all the samples.

Table 2 shows invention steel A (base metal, weld) and comparative steel G.
It is the result of performing the U vent test with a gap in the high temperature and high pressure of 320 degreeC (1000 hours) using (welded part) and H (base material). According to Table 2, for the steel A of the present invention, not only the base metal part That is, no cracks occurred at the welded portions (collection positions L, M, N). On the other hand, in the comparative steels G and H, cracked test pieces were already observed in the 500 hour test. From these results, it was proved that the steels of the present invention (A to F) have extremely excellent SCC resistance.

Table 3 shows the results of low strain rate tensile test of invention steel A and comparative steels G and I in hydrogenated water. As a result of the observation of the fracture surface, in the invention steel A, no SCC fracture surface was observed in both the base material and the welded member. On the other hand, 90% of the grain boundary type SCC was observed in Comparative Steel G and 20% of the grain boundary type SCC was observed in I.

Inferring from the above experimental results, it was confirmed that the invention steel is a material having extremely low SCC sensitivity even under the conditions of high temperature and high pressure water and hydrogenation.

FIG. 1 shows the tensile test results corresponding to the respective temperature levels of Invention Steel A and Comparative Steel I.

Inventive steel A retains high values in both tensile strength and 0.2% proof stress, and also has sufficiently high values at high temperatures, as compared with comparative steel I, which is a high Ni-based alloy.

Summarizing the above results, cobalt 5 by this activation
Despite the aim of developing a low nickel type stainless steel in which the amount of Ni in the material is limited in order to suppress the occurrence of No. 8, as described above, the SCC resistance in high temperature water is excellent and the high temperature strength is also high. There is a guideline for manufacturing a sufficiently large core material. This is largely due to the fact that the heat treatment conditions suitable for the steel of the present invention have been established, as well as the limitation on the chemical composition.

The effects of the reactor core equipment manufactured using the steels (A and E) of the present invention will be described with reference to FIGS. 2 and 3.

FIG. 2 is a partial perspective view of a core shell of a boiling water reactor manufactured using Invention Steel A.

In the configuration of FIG. 2, 1 is a shroud body, 2 is an upper shroud, 3 is a lower shroud, 4 is an upper guide,
5 is a core support plate. As described above, both the base metal portion and the welded portion have good SCC resistance and mechanical strength that is better than that of the comparative steel (conventional example).
While operating as an actual machine, it has achieved the expected effective result of reducing the radioactive contamination of cooling water.

Further, FIG. 3 is a partially cut perspective view of a control rod for a boiling water nuclear reactor manufactured by using the invention steel E. In the configuration of FIG. 3, 11 is a control rod cladding tube, 12 is a control rod sheath, 13 is a control rod tie rod, and 14 is a control rod handle.

Since the steel E of the present invention has a low content of P and S, even if it is subjected to high neutron irradiation, segregation of impurities does not occur. Since it has high strength and good elongation (%), it is possible to reduce the thickness of the plate and the pipe, which makes it possible to reduce the weight of the equipment and contribute to the improvement of earthquake resistance. Further, since the Ni content is reduced from 8.75% to 1.0% as compared with the conventional example (Comparative Steel G), the amount of cobalt 58 produced after irradiation can be reduced to about 1/9. did it.

With this effect, the dose rate of the actual equipment can be significantly reduced, and the amount of cobalt 58 adhering to the fuel cladding tube can be suppressed, so that it is possible to reduce the environmental radioactivity during the fuel reprocessing. Natsuta.

Although the above embodiments have described the core structure of a boiling water reactor, the steel of the present invention can of course be applied to dryers, steam separators and other internal reactor structures.

Further, it is also applicable to a pressurized water reactor, a pressure collecting pipe type heavy water reactor, or a natural circulation small reactor.

Needless to say, the above effects can be utilized even when applied to these furnaces.

〔The invention's effect〕

According to the present invention, the Ni content is greatly reduced as compared with the conventional austenitic stainless steel (comparative steel), and the steel of the present invention that has been appropriately heat-treated is used to manufacture core equipment, By using it, the occurrence of stress corrosion cracking can be prevented and the amount of cobalt 58 produced can be greatly reduced, so the radioactivity around the plant can be reduced, and the radiation exposure of employees during regular inspections can be reduced. Can contribute to.

In short, the minimum required N for reactor core equipment materials
By using a low nickel type stainless steel whose i content is suppressed and subjecting it to suitable heat treatment, Ni
It is possible to provide a stainless steel for nuclear reactor core equipment, which suppresses the elution of cobalt and suppress the generation of cobalt 58 due to activation, and a manufacturing method thereof.

[Brief description of drawings]

FIG. 1 is a comparison diagram of the mechanical properties of the invention steel and the comparative steel, and FIG.
The drawing shows a partial perspective view of a core shell made of the present invention steel (A), FIG. 3 is a partially cut perspective view of a control rod made of the present invention steel (E), and FIG. A) Quenching temperature and hardness / absorption energy relationship diagram, FIG. 5 is the tempering temperature and hardness / absorption energy relationship diagram of the same as above, and FIG. 6 is (C + N) content in steel and welding crack Correlation diagram, FIG. 7 is an explanatory view of taking out the stress corrosion cracking test piece from the welded joint member, FIG.
The figure is a U-bend test piece shape diagram with a gap, and FIG. 9 is a low strain rate tensile test piece shape figure. 1 ... Shroud trunk, 2 ... Upper shroud, 3 ...
Lower shell, 4 ... Upper guide, 5 ... Core support plate, 11 ... Control rod cladding tube, 12 ... Control rod sheath, 1
3 ... Control rod tie rod, 14 ... Control rod handle.

 ─────────────────────────────────────────────────── ─── Continuation of the front page (72) Takashi Saito Inventor Takashi Saito 3-1-1, Saiwaicho, Hitachi-shi, Ibaraki Hitachi Ltd. Hitachi factory (56) References JP-A-62-107047 (JP, A) JP 62-120464 (JP, A)

Claims (2)

[Claims] Claim: What is claimed is: 1. A material used to manufacture nuclear reactor core equipment, comprising Fe as a main component, and containing C, Si, Mn, P, S, Ni and C.
In an austenitic stainless steel containing r, Mo, Cu, Al and impurities, Ni = 1 to 6%, C <0.03%, Cr = 12 to 20%
In addition, N is added so that C + N <0.05%, and passes the U-bend corrosion test with a gap for 1000 hours in high-temperature high-pressure water. Stainless steel for reactor core equipment. 2. C, Si, Mn, P, S, Ni, C containing Fe as a main component, which is used for manufacturing nuclear reactor core equipment.
In the method for producing austenitic stainless steel containing r, Mo, Cu, Al and impurities, Ni = 1 to 6%, C <0.03%, Cr = 12 to 20%
Then, N was added so that C + N <0.05% and melt casting was performed. The cast product was heated at 950 ± 10 ° C. for 1 to 3 hours and then rapidly cooled.
A method for producing a stainless steel for nuclear reactor core equipment, comprising performing a tempering heat treatment at 00 ± 10 ° C. for 3 to 10 hours.