As discussed in earlier chapters, ductility (or toughness) is an important prop­erty of any structural material or any other types of materials in load-bearing nuclear reactor components. Radiation hardening generally leads to radiation embrittlement and occurs in a wide range of materials. However, radiation embrittlement in BCC metals/alloys (such as ferritic and ferritic-martensitic steels) that exhibit ductile-brittle transition temperature (DBTT) refers to an increased DBTT along with decreased upper shelf energy and decreased slope

in the ductile-brittle transition region, as shown in Figure 6.31a. Yield strength decreases with temperature, but fracture stress is roughly temperature indepen­dent. Irradiation causes an increase in the yield stress shifting the point at which fracture stress and flow stress curves intersect at higher temperature, thus raising the DBTT. Figure 6.31b is known as Davidenkov’s Diagram. It is to be noted that in BCC metals the yield stress increases rapidly as temperature decreases, while in FCC and HCP metals the decrease is not that rapid, which results in DBTT occurring at very low temperatures and thus ofless significance (Section 5.1). The effect of irradiation on the use is believed to be due to a reduction in strain hardening and increase in flow localization, leading to lower ductility (also dislocation channeling has an effect).

Earlier we have underscored the importance of having a decreased DBTT for structural applications. This is also applicable for reactor pressure vessel (RPV) steels, and this section is devoted to understanding the origin and implications of radiation embrittlement in RPV steels. Reactor pressure vessel is an integral part of a nuclear power reactor and it is considered a life-limiting reactor com­ponent, which means that the RPV is highly unlikely and extremely difficult (cost-prohibitive) to replace during the operational life of the reactor. Hence, the same RPV stays in place for the entire operational life of the reactor. Currently, many utilities are applying for relicensing their reactors for another 20-30 years as their original design life comes to an end. RPV surveillance pro­gram has facilitated the understanding of radiation embrittlement in RPV steels; however, understanding the behavior over very long period is still evolv­ing. The relicensing of the commercial power reactors beyond their original design lives makes the understanding of radiation embrittlement in RPV steels even more important. Understanding the role of late blooming phases in the

radiation embrittlement of reactor pressure vessel steels will need more sus­tained studies in the wake of LWR sustainability efforts.

Generally, low-alloy ferritic steels (A508 and A533B grades) are used in the RPVs of the light water reactors. The RPV steel in a PWR would have to withstand irradiation temperature of 240-290 ° C and fast neutron fluence of <2 x 1024 nm~2 over a time period of decades, leading to substantial changes in microstructure: (i) formation of radiation defects, (ii) phase transformations accompanied by the generation of various precipitate populations, (iii) formation of impurity-vacancy clusters (like copper-vacancy ones), (iv) intergranular segregation of phosphorus and other impurities, and (v) segregation of phosphorus to interfaces between sec­ondary phases and matrix (i. e., intragranular segregation). These are the likely rea­sons for radiation embrittlement as observed in the RPV steels.