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|Title:||SCC Properties and Oxidation Behaviour of Austenitic Alloys at Supercritical Water Conditions|
|Authors:||PENTTILÄ Sami; TOIVONEN Aki; HEIKINHEIMO Liisa; NOVOTNY Radek|
|Citation:||Proceedings of the 4th International Symposium on Supercritical Water-Cooled Reactors p. Paper No. 60 (1-16)|
|JRC Publication N°:||JRC52518|
|Type:||Contributions to Conferences|
|Abstract:||The goal of this paper is to determine the SCC susceptibility of candidate materials for SCWR incore applications. Results of SCC susceptibility at 500oC and 650oC on five dandidate materials are presented. Within the FP6 program "HPLWR Phase 2"-project (High Performance Light Water Reactor) general corrosion tests (i.e. oxidation rate tests) have been performed on several iron and nickel based alloys at 400oC to 650oC in supercritical water under the pressure of 25 MPa. The oxygen concentration of the inlet water was 0-150 ppb in all tests. The oxidation behaviour was studied using weight gain measurements, scanning electron microscopy in connection with energy dispersive spectorscopy (SEM and EDS, respectively) and X-ray diffractometry (XRD). Also, stress corrosion cracking (SCC) susceptibilities of selected austenitic stainless steels (i.e. 316NG, 1.4970, 347H and an experimental creep resistant stell GBA4) and a high chromium ODS (Oxide Dispersion Strengthened) alloy (i.e. PM2000) were studied in supercritical water (SCW) at 500oC and 650oC. The SCC tests were slow strain rate tests (SSRT) performed using a stop motor controlled loading device. The samples were strained with a nominal rate of 3x10 -7 s-1. Ferritic-martensitic (F/M) stells containing chromium have generally good resistance to stress corrosion cracking, but they suffer from fast oxidation in the SCW. Austenitic stainless steels and Ni-based alloys have better oxidation resistance and relatively good creep resistance but, on the other hand, are more susceptible to stress corrosion cracking than ferriticmartensitic steels. SSRT test showed that 316NG, 1.4970, 347H and PM2000 are not susceptible to SCC at 500oC based on fracture surface examination, but the experimental steel BGA4 showed a considerable susceptibility to intergranular SCC. Generally, at 650oC, the austenitic stainless stells were observed to be SCC susceptible, whihc corresponds well with the data reported in literature. The high chromium ODS steel PM2000 was SCC resistant at both test temperatures. Alloys with a high nickel content were considered for the SCC studies because Ni has a strong effect on neutronics of the reactor core. Therefore, the present candidate materials for the core internals are austenitic stainless steels and high chromium ODS alloys.|
|JRC Institute:||Institute for Energy and Transport|
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