Supercritical Water Reactor Review Meeting. Materials Issues

Transcription

1 Supercritical Water Reactor Review Meeting Materials Issues Bill Corwin, Louis Mansur, Randy Nanstad, Arthur Rowcliffe, Bob Swindeman, Peter Tortorelli, Dane Wilson, Ian Wright Oak Ridge National Laboratory Madison, Wisconsin April 30, 2003

2 ORNL Approach Examine the components and their required duty reactor internals and external circuit nominal stress level environment (thermal, radiological, and corrosion aspects) duty cycle issues abnormal conditions Review the range of materials available analogy to existing systems new materials Identify applicable information from other technologies (e.g., coal-fired steam generators) Flag SCWCR-specific issues Compile a report (5/31)

3 1. Candidate Materials for Reactor Internals: General Comments Zirconium alloys cannot be used because of their poor high-temperature corrosion resistance and mechanical strength Material austenitic ferritic/ Steels martensitic ODS High-Ni alloys Corrosion Resistance Fair? Properties Irradiation Stability Fair Fair (mech properties) (swelling) Fair Considerations that were used in our initial materials choices for SCWCR: peak temperatures in PWRS: C peak doses are prototypic for 40 yr life in LWR HT Mechanical Properties Fair Cost Low Low V. high High Typical Grade 316LN 9Cr alloys 12Cr-1Mo MA ; 690; 718 assumed peak dose for bolting is same as for baffle bolts and formers in PWR; water rod boxes same as core baffle plate in PWR

4 CR guide tubes Core barrel Pressure vessel Bolting Suggested Candidate Reactor Internal Materials Component Fuel Cladding Spacer grids/wire wrap Water rod boxes Fuel assembly ducts Lower core plate Upper support plate T C (inner) (outer) (inner) (outer) (upper) 500 (lower) (inner) 280 (outer) 280 (hot nozzle: 500) 370 Normal Conditions Peak dose dpa (?) 80?? Loads FGs & FCI secure fuel pins DP<0.1 MPa 700 for <30 sec DP<0.1 MPa 700 for <30 sec supports the core supports the CR guide tubes no load DP<0.1 MPa P = 25 MPa Abnormal Conditions T C 840 for <30 sec 840 for <30 sec N/A 600 for <30 sec (lower side) N/A N/A design P = 27.5 MPa Current LWR Materials PWR Zircaloy 4 Zircaloy 4 Inconel 718 type 304 SA 533-B-1 SA 503-2/3 type 316/CW 1 : T-91 (9Cr-1Mo-V); A-21 (9Cr-TiC); NF616 (9Cr); 9Cr-2WVTa; HCM12A (12Cr) BWR Zircaloy 2 Inconel X750 Zircaloy 2 type 304 Zircaloy 4 type 304 type 304 SA 533-B-1 SA 503-2/3 Candidate SCWCR Materials 1 low-swelling SS 2,3 low-swelling SS low-swelling SS Inconel 718 Inconel 625 low-swelling SS 2,3 Inconel 718 Inconel 625 low-swelling SS low-swelling SS A (A533-B-1) A-508-4N 3Cr-3WV type 316/CW Inconel 718 Inconel U.S. existing low-swelling stainless steels: D-9 (14.5 Cr-14.5Ni-2Mo-Ti); PNC (D-9 mod w/p) DEPARTMENT OF ENERGY 3 OAK new stainless RIDGE NATIONAL steels: 316FR; LABORATORY HT-UPS (± PNC); AL-6XN (20Cr-24Ni-6Mo-0.2Cu-0.2N)

5 Other Materials Issues for Reactor Internals 1. Ferritic/Martensitic vs Austenitic Steels F-M s have excellent swelling resistance, but are potentially sensitive to embrittlement some austenitic compositions have shown swelling resistance to dpa at peak swelling T 2. Use of ODS alloy MA-956 (Fe-Cr-Al-Y 2 O 3 ) as Cladding alloy was designed for creep strength at T > 1000 C essentially no data at low temperatures 475 C/885 F embrittlement? (ferritics typically susceptible to a-cr pptation at Cr levels > 13%) intended for compatibility with high-temperature gaseous environments forms an alumina scale (a-al 2 O 3 at T>1000 C) data on electrochemical behavior are sparse known to be susceptible to polythionic acid attack ( down-time corrosion ) ODS alloy MA-957 (Fe-Cr-Y 2 O 3 ) would be a better choice (if available)

6 2. External Circuit applicability of experience from coal-fired SC steam plants Combustor internals -- tubing subjected to heat flux from convection (gas) and radiation Main steam piping -- no pressure vessel penetrations Main steam valves (equivalent to MSIV) -- need to define expected duty Turbine -- nominally similar HP inlet conditions of T, P LP outlet conditions? Condenser-feedwater circuit -- nominally similar? no pressure vessel penetrations

7 Relevant Corrosion Issues for Water-Touched Surfaces Coal-Fired Boiler Coolant Water-treatment chemistry Acid phosphate corrosion Caustic gouging Chemical cleaning damage/pitting Nuclear Reactor Coolant Intergranular stress corrosion cracking (IGCC) Irradiation-assisted stress corrosion cracking (IASCC) Corrosion fatigue Classic aqueous corrosion Stress corrosion cracking Corrosion fatigue Hydrogen damage Fluid chemistry issues in nuclear units are significantly more complicated

8 ASME specification SA-178C SA-192 SA-210 A1 SA-209 T1 SA-209 T1a SA-213 T11 SA-213 T22 SA H SA H SA H Pressure-Boundary Materials Used in Coal-Fired SC Steam Generators 17/19Cr-9/12Ni-Ti typical use temperatures Alloy usage in coal-fired plants is very conservative. Only a limited number of alloys is used, and these are governed by the relevant ASME Boiler and Pressure Vessel Code requirements. C-Mo steel 2.25Cr-1Mo Alloy Carbon steel 1.25Cr-0.5Mo 18/20Cr-8/12Ni 17/19Cr-9/13Ni-NbTa Maximum Use Temperature ( C/ F) ASME* OEM-1 OEM-2 OEM-3 538/ / / / / / / / / / / / / / / / / / / / / / / / / / /1300 *Based on allowable strength with no corrosion considerations. Others based on experience/corrosion 454/ / / / / /1500

9 2a. Steam Turbine Issues There is experience in handling steam to 582 C/30.5 MPa (Avedøre) Example of inlet/outlet conditions (EPRI Advanced Plant): Turbine Stage VHP HP IP LP T C P, MPa Inlet T F P, psi T C P, MPa Outlet T F P, psi Feedwater to boiler (following feedwater heaters): 316 C, 35.9 MPa (600 F, 5,200 psi) For SCWCR at nominal operation: fluid inlet to reactor is at 280 C, 25 MPa (536 F, 3,626 psi) fluid outlet is at 510 C, 25 MPa (950 F, 3,626 psi)

10 Candidate Alloys-External Circuit & Turbine Component Main steam lines (nom: 510 C, 25 MPa/ 3,626 psi, 950 F; max: 600 C/1112 F) Castings Bolting Blades (nominal:510/950; max: 600/1112) Feedwater circuit (nominal: 280/536) Pumps (nom: 280 C, 25 MPa/ 536 F, 3,626 psi) Alloy P-22 P-91 P-92 P-122 1%Cr steels 12%Cr steels T-91 T-92 IN-718 Nimonic 80A type 422 IN-718 Nimonic 90 M252 (etc) T-11 T-22 T-23 CF3C CF3M CF8C CF8C+ T max C/ F 580/ / / /1200? 550/ / / / / / / / / / / / / / / / /1472 oxidation rate increases rapidly at higher Ts PWHT issues new alloys, few data industry standard; oxdtn rate increases rapidly at higher Ts industry standard (Europe) PWHT issues not heat-treated Issues based on strength based on strength (not heat-treated) based on strength based on strength based on corrosion in coal-fired units no PWHT needed (cast 304L--improved creep strength) (cast 316L--improved creep strength) (cast 347H--improved creep strength) ORNL development (PJM)

11 Specific Questions for SCWCR Operation Compatibility of reactor vessel internals with SC fluid does fossil water treatment experience have any relevance to nuclear SC steam generator conditions? Pressure vessel penetrations by main steam pipe some similarity to PFBC vessel penetration issues Sub-to-supercritical transformation: start-up and shut-down procedures? (fossil plants use turbine by-pass and valve coordination) how will load following be accomplished? Is transport of irradiated material into external circuit likely? exfoliation of oxide scales from steam lines? condensation in latter stages of turbine? Expected lifetime is 60 years what service intervals are envisioned? which components can be subjected to maintenance?