RADIATION DAMAGE

8.5 RADIATION DAMAGE

Radiation damage is the change of properties of a material caused by exposure to ionizing radiation, such as X - rays, gamma rays, neutrons, heavy - particle radia- tion, or fi ssion fragments in nuclear fuel material [65] . Metals exposed to intense radiation in the form of neutrons or other energetic particles undergo lattice changes resembling in many respects those produced by severe cold work. The porous structure produced by dealloying has been described as similar to that found in irradiated materials [66] . Lattice vacancies, interstitial atoms, and dislo- cations are produced, and these increase the diffusion rate of specifi c impurities or alloyed components. In copper and nickel at room temperature, radiation damage leads to increased hardness, caused by clusters of interstitials and vacan- cies [66] . During radiation, a local temperature rise, called “ temperature spike, ” may occur. There are two kinds of spikes: thermal spikes , in which few or no atoms leave their lattice sites, and displacement spikes , in which many atoms move into

interstitial positions. Except for chemicals produced in the environment by radiation, such as HNO 3 and H 2 O 2 , which have a secondary effect on corrosion, or formation of localized displacement spikes during radiation, the effect of radiation may be expected to parallel that of cold work. That is, metals for which the corrosion rate is controlled by oxygen diffusion should suffer no marked change in rate after irradiation. In acids, on the other hand, irradiated steel (but not pure iron) would presumably have a greater increase in rate than would irradiated nickel, which is less sensitive to cold working.

Austenitic stainless steels often become more sensitive to S.C.C. after cold working; on this basis, they might be expected to become more sensitive after irradiation. Indeed, irradiation - assisted stress - corrosion cracking (I.A.S.C.C.) is a cause of failure of core components in both boiling water reactors (BWR) and pressurized water reactors (PWR), and it has been observed, as intergranular S.C.C., in austenitic stainless steels and nickel - base alloys [67] . The complexities of S.C.C. are, in I.A.S.C.C., compounded by the effects of radiation on microstruc- ture, microchemistry, and deformation behavior of the material as well as on the chemistry and electrochemistry of the solution. I.A.S.C.C. of austenitic stainless steels is related to chromium depletion along grain boundaries. In addition to the grain - boundary chromium depletion that can result from sensitization during

CORROSION FATIGUE

welding (see Figs. 19.2 and Fig. 19.3 , Section 19.2.3.1 ), chromium depletion in the grain boundary can also occur because of neutron irradiation, further reducing resistance to S.C.C. [68, 69] . One materials solution to mitigate I.G.S.C.C. in nuclear reactors is to use nuclear - grade stainless steels, such as type 316NG and type 304NG, which have a maximum carbon content of 0.020% and a nitrogen content of 0.060 to 0.100% to maintain strength at the lower carbon content compared to the non - nuclear grades (see Table 19.2 , Section 19.2.2 ) [69] . Niobium -

stabilized, type 347 stainless steel, with low carbon called 347NG, has also been used successfully [70] . Of course, good corrosion design is also essential — for example, a design without crevices, which can provide sites for localized corrosion (see Sections 2.3 and 19.2 .).

The effect of irradiation on corrosion of some uranium alloys is consider- able. For example, a 3% Cb – U alloy having moderate resistance to water at 260 ° C disintegrated within 1 h after irradiation. Furthermore, the corrosion of a zirconium alloy (Zircaloy - 2, see Section 26.2 ) at 250 ° C in dilute uranyl sulfate

solution containing small amounts of H 2 SO 4 and CuSO 4 was very much increased by reactor irradiation [71] . In a review of the subject, Cox [72] stated that both fast neutron irradiation and presence of dissolved oxygen or an oxidizing elec- trolyte must be present simultaneously for any observed acceleration of corro- sion to occur in high - temperature water. Accelerated corrosion of Zircaloys induced by irradiation is not observed above 400 ° C (750 ° F). The effects have been explained in terms of changes in the physical properties of the protective oxide fi lm.