Metal treatment – Stock – Ferrous
Reexamination Certificate
1999-04-19
2001-06-12
Yee, Deborah (Department: 1742)
Metal treatment
Stock
Ferrous
C148S327000, C148S607000, C148S608000
Reexamination Certificate
active
06245163
ABSTRACT:
BACKGROUND OF THE INVENTION
1. Field of the Invention
The present invention relates to a thermal treated austenitic stainless steel having excellent resistance to neutron-irradiation-induced deterioration. The thermal treated austenitic steel is used, for example, as a structural member inside a light water reactor type nuclear power plant.
2. Description of the Related Art
Austenitic stainless steels, such as SUS 304 and SUS 316, which have conventionally been used for a structural member (bolt, plate and the like) inside a light water reactor type nuclear power plant, tend to lack Cr or have concentrated Ni, Si, P, S and the like in its grain boundary after long years of use and subjected to neutron irradiation of 1×10
21
n/cm
2
(E>1 MeV) or greater. Under the stress of a high load, the austenitic stainless steel tends to exhibit stress corrosion cracking (SCC) under use in a light water reactor. Such a phenomenon is called “irradiation-affected stress corrosion cracking” (IASCC). Although there is a strong demand for the development of a material having low IASCC sensitivity, or in other words, having excellent resistance to neutron-irradiation-induced deterioration, no such material has yet been developed.
Austenitic stainless steels such as SUS 304 and SUS 316 have been used as structural material inside a light water atomic power reactor. When such materials have been in use for long periods of time and subjected to neutron irradiation of 1×10
21
n/cm
2
(E>1 MeV) or greater, an undesirable change is observed in the concentration of certain elements in the vicinity of the grain boundary which was not present or was present only to a slight extent prior to use. In other words, a lack of Cr and Mo or enrichment of elements such as Ni, Si, P and S in the vicinity of the grain boundary is observed. This phenomenon is called “irradiation-induced segregation”. It is known that in this segregated state, the presence of a high load stress or residual stress tends to cause stress corrosion cracking (irradiation affected stress corrosion cracking: IASCC) in water of high temperature and high pressure, the neutron irradiated environment of a light water reactor.
The present inventors had developed and previously proposed a Ni-rich austenitic stainless steel as a material having excellent resistance to neutron-irradiation-induced deterioration. The Ni-rich stainless steel of a specific composition was treated thermally to optimize the crystalline form of the alloy and then subjecting the resultant steel to post processing. (Japanese Patent Application Laid-Open No. 9-125205).
SUMMARY OF THE INVENTION
An object of the present invention is to provide a structural material which has been used conventionally, such as SUS 304, SUS 316 or SUS 310S specified in JIS (Japanese Industrial Standards) as a base alloy that is resistant to neutron-irradiation-induced deterioration, without having to use a stainless steel having a high Ni content, and yet does not exhibit stress corrosion cracking (SCC) in water of high temperature and high pressure, the environment of a light water reactor.
The present inventors have conducted various investigations on the properties of austenitic stainless steel. Employing the method of measuring the value of the intergranular segregation of a neutron irradiated material as described by S. Dumbill and W. Hanks (Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 521(1993)), the concentration changes of Cr and Ni in the grain boundary were calculated by the present inventors. The concentration changes were compared with the test results of SCC of neutron-irradiated SUS 304 and SUS 316 collected by the present inventors and shown in FIG.
1
. The data shows that IASCC occurs when the concentration of Cr is reduced to less than 15% and that of Ni increased to more than 20% in the grain boundary after neutron irradiation. The cross hatched portion of
FIG. 1
illustrates the region of SCC generation.
The present inventors hypothesized that the phenomenon of IASCC occurs because the concentration of elements in the grain boundary approaches to that of the composition of Alloy 600 (NCF600 of JIS). Specifically, neutron irradiation reduces the Cr concentration and increases the Ni concentration in the grain boundary making the composition therein proximate that of Alloy 600 (non-irradiated material; Ni≧72%, Cr=14~17%). It had been frequently observed that with Alloy 600, stress corrosion cracking (PWSCC: stress corrosion cracking which occurs in a primary water) occurs in water of high temperature and high pressure. However, up to the present, the mechanism of PWSCC has not been elucidated in detail.
It has been known that in conventional Ni-based alloys, Incone 1750 (NCF750 of JIS) or Alloy 690 (NCF690 of JIS), the grain boundaries are strengthened with improved PWSCC resistance by subjecting it to special thermal aging treatment under specific conditions. The treatment caused precipitation, in the grain boundary, of M
23
C
6
(a carbide having mainly Cr as M) to matching that of the matrix phase. The present inventors have found that when the special thermal treatment employed for a Ni-based alloy is applied to the conventional SUS 304, SUS 316 or SUS 310S, the grain boundary can be reinforced and SCC resistance improved by precipitating M
23
C
6
matching that of the matrix phase in the grain boundary, even if neutron irradiation lowers the Cr concentration and raises the Ni concentration of the composition in the vicinity of the grain boundary.
Based on these findings, the present inventors proceeded to further investigate and develop the present invention, wherein SUS 304 or SUS 316 was employed as a base alloy and applying a combination of solid solution treatment under specific conditions, aging thermal treatment to optimize the crystalline form in the alloy and post-processing (cold working) treatment.
The present invention provides an austenitic stainless steel having resistance to neutron-irradiation-induced deterioration obtained by subjecting a stainless steel consisting of not more than 0.08% by weight of C, not more than 2.0% by weight of Mn, not more than 1.5% by weight of Si, not more than 0.045% by weight of P, not more than 0.030% by weight of S, 8.0 to 22.0% of by weight Ni, 16.0 to 26.0% of by weight Cr and the balance of Fe, to thermal solid solution treatment at 1,000° to 1,180° C. and then to aging treatment at 600° to 750° C.
Optionally, the austenitic stainless steel according to the present invention can be obtained by further subjecting said stainless steel to cold working treatment up to 30% of the cross-sectional area of the material subjected to thermal solid solution treatment and aging treatment.
The stainless steel used in the present invention may contain 3.0% by weight or less of Mo. For example, the stainless steel may be SUS 316 specified in JIS. When SUS 316 is used, the temperature range of said thermal solid solution treatment is 1,000° C. to 1,150° C.
Further, the stainless steel may be SUS 304 specified in JIS. When SUS 304 is used, the temperature range of said thermal solid solution treatment is 1,000° C. to 1,150° C.
Furthermore, the stainless steel may be SUS 310S specified in JIS. When SUS 310S is used, the temperature range of said thermal solid solution treatment is 1,030° C. to 1,180° C.
REFERENCES:
patent: 5976275 (1999-11-01), Yonezawa et al.
patent: 5987088 (1999-11-01), Aono et al.
patent: 52-52116 (1977-04-01), None
patent: 093001319 (1993-01-01), None
Fujimoto Koji
Iwamura Toshihiko
Yonezawa Toshio
Mitsubishi Heavy Industries Ltd.
Morgan & Finnegan , LLP
Yee Deborah
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