The Mutual Influence of Materials and Thermal-hydraulics on Design of SCWR Review of Results of the Project HPLWR Phase 2

Transcription

1 Institute of Nuclear Technology and Energy Systems The Mutual Influence of Materials and Thermal-hydraulics on Design of SCWR Review of Results of the Project HPLWR Phase 2 J. Starflinger, T. Schulenberg

2 HPLWR High Performance Light Water Reactor 5th Framework Programme of the EU Design Target Data: Operational pressure: Core mass flow: Power output: 25 MPa 1160 kg/s 1000 MWe Constraints: Average core exit temp.: 500 C Max. cladding surface temp.: 625 C Max. linear heat rate: 39 kw/m AREVA NP, 2005 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/2016 2

3 Hot Channel Form Factor Analysis of the Core Definition F h h max av h h max av Maximum enthalpy rise in the Hot Channel Average enthalpy rise in the core Hot Channel by definition is the channel, in which all uncertainties, nonhomogeneities and allowances sum up, leading to the highest enthalpy rise of the entire core under normal operation conditions! Very conservative, provides a very high safety margin IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/2016 3

4 Design Targets of Hot Channel Factors Hot Channel Factor axial radial Key Parameters Form factors for power profiles Fuel enrichment and distribution, water density distribution, reflector design and properties, fuel and control rod pattern, burn-up, burnable poisons, Radial peaking factor 1.25 Local peaking factor inside FA 1.15 Axial power factor 1.6 Uncertainties 1.2 Material properties of coolant and claddings, physical modelling, hydraulic modelling, heat transfer coefficient, geometry tolerances Allowances 1.15 Power control, flow control, pressure control, inlet temperature control Total Schulenberg, KIT, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/2016 5

5 Hot Channel Form Factor Analysis of the Core One-Pass Core Designed for 500 C core outlet temperature Coolant average conditions Average Heinecke, AREVA, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/2016 6

6 Hot Channel Form Factor Analysis of the Core One-Pass Core Hot fuel assembly ( 1.25) + Assembly Power Average Heinecke, AREVA, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/2016 7

7 Hot Channel Form Factor Analysis of the Core One-Pass Core Hot rod ( = 1.44 ) + Rod Power + Assembly Power Average Heinecke, AREVA, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/2016 8

8 Hot Channel Form Factor Analysis of the Core One-Pass Core + Assembly Power + Uncertainty + Rod Power Hot rod + uncertainty ( = 1.73 ) Average IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/2016 9

9 Hot Channel Form Factor Analysis of the Core One-Pass Core + Operation + Uncertainty + Rod Power + Assembly Power Average Hot rod + uncertainty + operation ( = 1.98 ) Coolant temperature 1200 C Cladding surface temperature > 1200 C Heinecke, AREVA, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

10 Hot Channel Form Factor Analysis of the Core Consequences Simple Hot-Channel analysis revealed the unfeasibility of single-pass core design. No material available. Idea from T. Schulenberg, KIT: Propose a Three-pass core with intermediate mixing in special mixing chambers. One key-issue of a feasible core design is mixing! not to overheat the cladding surface temperature avoid hot streaks from one assembly to another and hot-spots on the cladding surface IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

11 Enthalpy [kj/kg] Three Pass Core Design Proposal for a HPLWR Strategy to overcome hotchannel issue: Power ratio of the core zones 4 : 2 : 1 Heat-up in steps with Intermediate mixing of the coolant Mixing 2000 Mixing 1500 hot channel average Schulenberg, Evaporator Superheater 1 Superheater 2 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

12 Temperatures [ C] Three Pass Core Design Proposal for a HPLWR cladding hot channel average Evaporator Superheater 1 Superheater 2 Schulenberg, 2006 A 3-Pass coolant flow in the core allows 500 C average core exit temperature with 625 C cladding temperature IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

13 Core Arrangement Superheater 1: 52 Clusters Downward Flow Evaporator: 52 Clusters Upward Flow Superheater 2: 52 Clusters, Upward Flow Köhly, 2010 IAEA - Technical Meeting on Heat transfer, Thermal-hydraulics, System Design for SCWR 22/8//

14 HPLWR Flow Path Upper dome Moderator flow (50%) Inlet flow: 280 C 25 MPa 1179 kg/s Downcomer flow (50%) Downcomer Area Core flow (100%) IAEA - Technical Meeting on Heat transfer, Thermal-hydraulics, System Design for SCWR Köhly, /11/

15 Tasks for the HPLWR Partners Design Support Analyze the core, whether the power peaking factors will be met Neutronics: Simulate neutronics (BOC, EOC) for core and assembly-wise power distribution (input from materials and TH needed) Thermal-hydraulics: Suitable heat transfer correlation with an uncertainty of less than 25%, especially for fuel rod bundles with wire wraps as spacers. Materials & Water Chemistry Identify suitable materials for thick wall and thin wall components, especially for cladding. IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

16 Materials Partners: VTT, Finland, NRI Czech Republic, CEA, France, EKI, Hungary, JRC-IE, Petten (supporting) Test of 16 materials in autoclaves at different temperatures Investigation of general corrosion, stress-corrosion cracking and creep Special focus on thin wall materials IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

17 Details of the Assembly Design Concept 40 fuel pins with 8mm diameter cladding thickness 0.5 mm wire wraps as grid spacers wire diameter 1.34 mm assembly box with 3 mm thickness incl. thermal insulation moderator box with 2 mm thickness incl. thermal insulation Insulation material: ZrO 2 Moderator box Assembly box Wire wrap spacers Himmel, Köhly 2008 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

18 Fuel Rods with Wire Wrap Spacers Local Temperature Distribution bare rod with wire Temperature [ C] Lycklama, 2009 Temperatures > 670 C Bending (stresses, torque) Hot streaks (local corrosion) IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

19 Materials General Corrosion Results 400 C 500 C 650 C 650 C, HWC Sample Holder, VTT Heikinheimo, 2009 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

20 Effect of Surface Treatment Strong impact of cold work on corrosion Alloy 316L tube samples after 1000 h exposure under SCW conditions at 650 o C: Machined with blunt edge hard metal cutting tool as received surface finish with #600 emery paper surface finish with #1200 emery paper Heikinheimo, 2009 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

21 Oxide Thickness (mm) Materials Influence of high Cr content on oxidation 1000 after 600h at 650 C P91 P ODS (FZK) ODS (EU) 10 PM NG BGA4 800H IN 625 0, Cr(%) Data from VTT, JRC, UJV Rez IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

22 Materials Extrapolation of oxide thickness Oxide thickness on AISI 316NG vs. exposure time after exposure to supercritical water at 650 C. 50% cladding thickness 10% cladding thickness Toivonen, Pentillä, 2009 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

23 Deformation Analysis of the Box Initial temperature distribution Outlet: 518 C Inlet: 286 C 3 mm stainless steel plate filled with ZrO 2 Reis, 2008 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

24 Deformation Analysis of the Box High high temperatures Max. deflection 4 mm (high strain) Reis, 2008 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

25 Stress-strain curves at 500 C High yield strength (good for designers) Toivonen, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

26 Stress-strain curves at 650 C Reduced yield strength (concern for designers) Toivonen, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

27 Stress-Corrosion cracking Alloy Maximum stress (MPa) Strain to failure (%) TGSCC (y/n) IGSCC (y/n) Side cracks at the gauge surface (y/n) 500 C 650 C 500 C 650 C 500 C No No Yes, morphology not identified NA NA NA NA NA Interrupted at 330 Interrupted at 33 NA NA Yes, IG and TG Yes Yes No No No Yes, morphology not identified Toivonen, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

28 Stress-Corrosion cracking (cont d) Alloy Maximum stress (MPa) Strain to failure (%) TGSCC (y/n) IGSCC (y/n) Side cracks at the gauge surface (y/n) 650 C 500 C 650 C 500 C PM C Badly oxidized Badly oxidized Yes, morphology not identified Yes Yes Yes, IG NA NA NA NA NA 325 Interrupted at 50 No No No No No No Toivonen, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

29 Creep From the designers point of view Design rules for high temperature applications Creep strength (650 C, 250MPa) Rupture Tensile yield 1% strain SCW: 25 MPa, 100 ppb O 2, deionized water k<0,1µs/cm (inlet), water flow rate 2-3 ml/min Gas: Helium Toivonen, 2010 IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

30 Creep SCW vs. He-atmosphere SCW environment increases strain rate compared to He environment for 316NG and 347H (short duration tests, usually > 1000 hrs) Strong impact on design expected! IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

31 Creep SCW vs. He-atmosphere SCW environment increases strain rate compared to He environment for 316NG and 347H (short duration tests, usually > 1000 hrs) Strong impact on design expected! The reasons for the increased primary strain rate: Give me more money and I will find out why IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

32 Example from Stress Analysis Thick walled component Design pressure: MPa Design temperature: 280 to 500 C Material: Wall thickness Core base plate: m Bottom plate: m Separation wall: m Outer wall: m Weight: 24.9 t IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

33 Example from Stress Analysis Thick walled component 3-D 10-Node Tetrahedral structural + thermal solids: nodes elements IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

34 Results: Deformation (cold state) Max.: 0.7 mm IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

35 Results: Mises stress distribution (cold state) Max.: 54 MPa IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

36 Results: Temperature distribution (operational/steady state) 433 C 309 C IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

37 Results: Deformation (operational/steady state) Max.: 12.5 mm IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

38 Results: Mises stress distribution (operational/steady state) Max.: 603 MPa IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

39 Results: Mises stress distribution (operational/steady state) Max.: 603 MPa Allowed stress peaks: s < 2 x R p0, : YS (600 C) = 350 N/mm² => 700 N/mm² Calculated values: 603 N/mm² => below 700 N/mm² => ok! IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

40 Summary Main findings Materials, thermal hydraulics have a strong mutual interaction on design of SCWR. Thick walled components operating at max. 500 C No major structural problems with respect to corrosion (fossil plant technology). Thin walled components operating at max. 500 C: Corrosion problems to be avoided (high Cr steels?) at above 600 C: High corrosion rate with licensed low Ni alloys (especially fuel cladding) High impact on core design! Redesign necessary if no suitable materials will be found. IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

41 Conclusions Some personal thoughts Closer collaboration between the scientific fields should be established: Theory of McElligott and Laurien: Surface roughness plays a role in heat transfer (disturbing boundary layers) Wire wrap helpful to homogenize the surface temperature. However, hot streaks are visible (also close to the wire) -> Local oxidation? Neutronic aspects must be taken into account Water chemistry may decide, whether to build a BWR or a PWR-type SCWR In-pile experiments are the next step for SCWR materials selection! (for heat transfer: bundle experiments to be performed) IAEA - Technical Meeting on Materials and Chemistry for Supercritical Water Cooled Reactors 10/10/

42 Institute of Nuclear Technology and Energy Systems Thank you! Prof. Dr.-Ing. Jörg Starflinger phone +49 (0) fax +49 (0) Universität Stuttgart Institute of Nuclear Technology and Energy Systems Pfaffenwaldring Stuttgart Germany