[en] The effects of cold work and low-temperature sensitization heat treatment of non-sensitized austenitic stainless steels have been investigated and related to the cracking in nuclear power reactors. Types 304, 304L and 304LN developed martensite after 15% cold working. Heat treatment of these cold worked steels at 500 deg. C led to sensitization of grain boundaries and the matrix and a desensitization effect was seen in 11 days due to fast diffusion rate of chromium in martensite. Types 316L and 316LN did not develop martensite upon cold rolling due to its chemical composition suppressing the martensite transformation (due to deformation) temperature, hence these were not sensitized at 500 deg. C. The sensitization of the martensite phase was always accompanied by a hump in the reactivation current peak in the double loop electrochemical potentiokinetic reactivation test, thus providing a test to detect such sensitization. It was shown that bending does not produce martensite and therefore, is a better method to simulate weld heat affected zone. Bending and heating at 500 deg. C for 11 days led to fresh precipitation due to increased retained strain and desensitization of 304LN due to faster diffusion rate of chromium along dislocations. The as received or solution annealed 304 and 304LN with 0.15% nitrogen showed increased sensitization after heat treatment at 500 deg. C, indicating the presence of carbides/nitrides

[en] Present work deals with studies relating to passivation, pitting and intergrannular corrosion behaviour of borated stainless steel. In addition effect of heat treatment e.g. sensitisation on corrosion resistance of this alloys has also been examined

[en] Zircaloy-2 has shown acceptable corrosion rates as fuel tubes in PHWR environment when the impurity nitrogen is well within the specified limit (80 ppm). There are uncertainties on the overall corrosion behaviour if the nitrogen content exceeds this value. This paper summarizes some of the data on zircaloy-2 samples containing nitrogen levels ranging from 120-220 ppm. Standard 14 day steam corrosion tests (400 C, 1500 psi) carried out on these samples have shown weight gains between 34 to 39 mg/dm², just close to the acceptable limit of 38mg/dm², for standard zircaloy-2 material. Tests were then carried out in high temperature water at 350 C (saturation pressure) for periods upto 40 days. The weight gains after 40 days of exposure at 350 C ranged between 25 to 35 mg/dm². Comparing these values against the weight gain time of exposure plots for standard zircaloy-2 at 350 C and 316 C, and assuming an acceleration factor of max. 3.0 under PHWR conditions, it can be stated that these zircaloy samples will attain a weight gain of 35 mg/dm² after 180 days under reactor operating conditions (290-310 C). This would correspond to an oxide thickness of 2.2 μm. The data further suggests that for a stay-in-time of about 900 days, the oxide thickness may range between 10-12 μm, which appears to be an acceptable value for fuel tube application. (author). 2 tabs

[en] Because of their austenitic-ferritic microstructures, duplex stainless steels offer a good combination of mechanical and corrosion resistance properties. However, heat treatments can lower the mechanical strength of these stainless steels as well as render them susceptible to intergranular corrosion (IGC) and pitting corrosion. In this study, a low-carbon (0.02%) duplex stainless steel is subjected to various heat treatments at 450 to 950 C for 30 min to 10 h. The heat-treated samples than undergo ASTM IGC and pitting corrosion tests, and the results are correlated with the microstructures obtained after each heat treatment. In the absence of Cr₂₃C₆ precipitation, σ-phase precipitates render this duplex stainless steel susceptible to IGC and pitting corrosion. Even submicroscopic σ-phase precipitates are deleterious for IGC resistance. Longer-duration heat treatments (at 750 to 850 C) induce chromium diffusion to replenish the chromium-depleted regions around the σ-phase precipitates and improve IGC resistance; pitting resistance, however, is not fully restored. Various mechanisms of σ-phase formation are discussed to show that regions adjacent to σ-phase are depleted of chromium and molybdenum. The effect of chemical composition (pitting resistance equivalent) on the pitting resistance of various stainless steels is also noted

[en] The oxidation behaviour of zirconium-1Nb, zirconium-2.5Nb and zircaloy-2 has been studied in 400deg C steam for 72 hours at 20 MPa with proper control of oxygen as per the procedure described in G 2M of ASTM. The oxide coatings were observed to be uniform and lustrous black. In a separate test, the Zr - 1 Nb samples were exposed in an autoclave in the same conditions without the control of oxygen and the oxygen content was high in this test. These tests showed about 5 fold increase in oxidation rate. In separate standard tests as per ASTM G 2M, it was shown that pressure (up to 20 MPa) does not play any significant role while etching (pickling) does affect the oxidation rates of Zr-1Nb material. (author)

[en] This paper highlights that though the use of nitric acid grade stainless steels (NAG SS) has ensured freedom from intergranular corrosion (IGC) of welded components, other degradation issues have to be addressed to avoid corrosion related failures in plants that operate with nitric acid as the main process fluid. End grain corrosion is a major form of degradation that affects all grades of stainless steels including NAG stainless steels. Controlled solution annealing, laser surface remelting, weld overlay and welding of a material over the exposed end faces have been shown to be highly effective in avoiding end grain corrosion. Other factors that are of concern are the presence of halide ions in nitric acid streams and the influence of cold work in fabricated components on the corrosion behaviour of stainless steels. Implementation of these degradation control measures, related to fabrication and process chemistry, would ensure trouble free operation of these plants. (author)

[en] Owing to the low neutron absorption cross section coupled with desirable mechanical properties, zirconium based alloys find their major application in structural materials such as pressure tubes and fuel clads in nuclear reactors. Although they exhibit a better corrosion resistance than many other materials, oxidation at high temperature and pressure steam/water remains a major concern under service. Upcoming advanced reactors, being designed worldwide, are adopting partial boiling environments for increased power ratings. The results are expected to have a direct consequence on the selection of pressure tube and fuel clad materials in the upcoming advanced nuclear reactors that allow partial boiling

[en] Nitric acid is used as a process fluid in nuclear fuel reprocessing and waste management plants. Type 304L stainless steel is the most widely used structural material and is used for handling HNO₃ over a wide range of temperatures and concentrations. To safeguard against intergranular corrosion (IGC), nitric acid grade (NAG) of type 304L has been developed. This stainless steel does not undergo sensitization during welding but still remains prone to corrosion attack by end grain corrosion