SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions

Sami Penttilä, Aki Toivonen, Liisa Heikinheimo, Radek Novotny

    Research output: Contribution to conferenceConference articleScientific

    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.
    Original languageEnglish
    Number of pages16
    Publication statusPublished - 2009
    MoE publication typeNot Eligible
    Event4th International Symposium on Supercritical Water-Cooled Reactors
    - Heidelberg, Germany
    Duration: 8 Mar 200911 Mar 2009

    Conference

    Conference4th International Symposium on Supercritical Water-Cooled Reactors
    CountryGermany
    CityHeidelberg
    Period8/03/0911/03/09

    Fingerprint

    Stress corrosion cracking
    Oxidation
    Water
    Chromium
    Austenitic stainless steel
    Oxides
    Steel
    Strain rate
    Nickel
    Gain measurement
    Scanning electron microscopy
    Creep resistance
    Light water reactors
    Reactor cores
    Oxidation resistance
    Weighing
    X ray diffraction analysis
    Energy dispersive spectroscopy
    Creep
    Corrosion

    Keywords

    • supercritical water
    • stress corrosion cracking
    • austenitic alloys
    • ODS steel
    • SSRT

    Cite this

    Penttilä, S., Toivonen, A., Heikinheimo, L., & Novotny, R. (2009). SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions. Paper presented at 4th International Symposium on Supercritical Water-Cooled Reactors
    , Heidelberg, Germany.
    Penttilä, Sami ; Toivonen, Aki ; Heikinheimo, Liisa ; Novotny, Radek. / SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions. Paper presented at 4th International Symposium on Supercritical Water-Cooled Reactors
    , Heidelberg, Germany.16 p.
    @conference{9d838306b07f4985a5274d26993221e0,
    title = "SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions",
    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.",
    keywords = "supercritical water, stress corrosion cracking, austenitic alloys, ODS steel, SSRT",
    author = "Sami Penttil{\"a} and Aki Toivonen and Liisa Heikinheimo and Radek Novotny",
    note = "Project code: 6366 Project code: 20062 ; 4th International Symposium on Supercritical Water-Cooled Reactors<br/> ; Conference date: 08-03-2009 Through 11-03-2009",
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    language = "English",

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    Penttilä, S, Toivonen, A, Heikinheimo, L & Novotny, R 2009, 'SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions', Paper presented at 4th International Symposium on Supercritical Water-Cooled Reactors
    , Heidelberg, Germany, 8/03/09 - 11/03/09.

    SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions. / Penttilä, Sami; Toivonen, Aki; Heikinheimo, Liisa; Novotny, Radek.

    2009. Paper presented at 4th International Symposium on Supercritical Water-Cooled Reactors
    , Heidelberg, Germany.

    Research output: Contribution to conferenceConference articleScientific

    TY - CONF

    T1 - SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions

    AU - Penttilä, Sami

    AU - Toivonen, Aki

    AU - Heikinheimo, Liisa

    AU - Novotny, Radek

    N1 - Project code: 6366 Project code: 20062

    PY - 2009

    Y1 - 2009

    N2 - 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.

    AB - 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.

    KW - supercritical water

    KW - stress corrosion cracking

    KW - austenitic alloys

    KW - ODS steel

    KW - SSRT

    M3 - Conference article

    ER -

    Penttilä S, Toivonen A, Heikinheimo L, Novotny R. SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions. 2009. Paper presented at 4th International Symposium on Supercritical Water-Cooled Reactors
    , Heidelberg, Germany.