Super-Critical Water-cooled Reactor

Transcription

1 Super-Critical Water-cooled Reactor SCWR System Steering Committee Y.P. Huang (Chair, China) L. Leung (Co-Chair, Canada, Presenter) R. Novotny (EU, awaiting confirmation) A. Sedov (Russian Federation) H. Matsui (Japan) Slide 1

2 Contents Gen-IV SCWR System Members and Projects System Integration and Assessment Project Thermal-Hydraulics and Safety Project Materials and Chemistry Project Fuel Qualification Testing Project Contribution to Projects ISSCWR-7 and 11 th SCWR Information Exchange Meeting Extension of the SCWR System Arrangement Summary Slide 2

3 Gen-IV SCWR System Members and Projects SCWR System Arrangement (year of sign.) and Representatives Canada (2006) L. Leung (SSC Co-Chair), D. Brady China (2014) Y.P. Huang (SSC Chair), L.F. Zhang Euratom (2006) R. Novotny, M. Krykova Japan (2006) H. Matsui Russian Federation (2011) A. Sedov, A. Churkin Projects: System Integration and Assessment (provisional) Thermal-Hydraulics and Safety (signed 2009)» Signatories: Canada, Euratom, Japan» China s joining procedure in progress» Russian Federation expressed interest to join Materials and Chemistry (signed 2010)» Signatories: Canada, Euratom, Japan» China s joining procedure in progress Fuel Qualification Testing (provisional) Slide 3

4 System Integration & Assessment Project (Provisional) Canadian SCWR design concept with pressure tubes 336 vertical fuel channels 2500 MW thermal power 1200 MW electric power 625 o C core outlet temp. 48% efficiency Conceptual Design completed and assessed in 2015 Slide 4

5 System Integration & Assessment Project (Provisional) China SCWR design concept with pressure vessel-csr1000 Technical Parameters thermal power Value 2300 MW electric power ~1000 MWe Efficiency ~43% operating pressure design pressure reactor inlet temperature reactor outlet temperature reactor flow rate 25 MPa 27.5 MPa 280 C 500 C 4284 t/h (1190 kg/s) loop number 2 Design to be completed and assessed in 2018 Cycle coolant flow-path design lifetime direct once-through two-pass 60 years Slide 5

6 Thermal-Hydraulics and Safety Project Status Heat transfer data at supercritical pressures for rod bundles with prototypical spacer geometry have been obtained with Supercritical water Supercritical CO 2 Supercritical refrigerants These data can be used to validate codes and to improve prediction methods. A new joint benchmark exercise is being prepared to start in 2016 The Project Plan is being updated to capture potential contributions of Canada, Euratom and China from Slide 6

7 TH&S Project Achievements Bare rods 4-rod bundle with upward flow of SC water Wire-wrapped rods 3-rod bundle with upward flow of SC CO 2 Slide 7

8 TH&S Project - Joint Benchmark Exercise Organized by M. Rohde, Technical University at Delft Benchmark analytical tools against a unique set of experimental data on heat transfer to SCW of a 7-rod bundle (contributed by JAEA, Japan) Blind predictions by 10 organizations from EU and Canada Identified gaps for further improvements Slide 8

9 TH&S: Future contributions planned for the period from 2016 to 2020 Heat transfer to supercritical water in tubes, annuli, sub-channels and rod bundles (CA, CN, RF) Heat transfer to supercritical CO 2 and Freon in tubes, annuli and rod bundles; analysis of fluid-to-fluid scaling laws (CA, CN, EU, RF) Pressure loss of supercritical water flow in rod bundles (CN, RF) Test of rod cladding ballooning (RF) Blow-down experiments with supercritical water (CA, CN, RF) Flow instabilities (CA, CN, EU, RF) SCWR safety requirements and evaluation (CA, EU, CN, RF) System code development (CA, CN) CFD and turbulence modelling (CA, CN, EU, RF) Slide 9

10 Materials and Chemistry Project Achievements Completed commissioning of the out-of-pile supercritical water loop at CVR/Rez (EU) Completed documentation of results of round-robin corrosion tests (CA, EU, JP) Developed Materials Databases (CA, EU) Quantified effects of coatings and surface modification (CA, EU) Identified applicable commercial alloys for SCWR applications in terms of general corrosion, stress corrosion cracking susceptibility and structural integrity (CA, EU) Assessed physico-chemical properties of SCW on materials corrosion behavior and general corrosion mechanism in SCW (CA, EU) Initiated the development of reference electrodes and test facilities capable of working under in-situ reactor conditions (EU) Specified water chemistry control strategy and developed a water radiolysis model (CA) Slide 10

11 Materials and Chemistry Project Sample Results SSRT curves for 316L steel in air SSRT curves for HR3C steel in SCW SSRT curves for 310S and HR3C steels in SCW SSRT curves for 316L steel in SCW Slide 11

12 Modeling of radiolysis in Canadian SCWR Predicted concentrations of radiolysis products produced in the Canadian SCWR core Predicted effect of H 2 addition on the concentration of oxidizing species and comparison with measurements from Beloyarsk NPP in superheated steam Slide 12

13 M&C: Future contributions planned for the period from 2016 to 2020 Study of the effect of surface finish and water chemistry on corrosion behaviour in supercritical water (CA, CN, EU) An iron/iron oxide reference electrode development work for insitu corrosion monitoring in supercritical water (EU) Creep tests of SS 347H and SS 310S (CA, CN, EU) Stress corrosion cracking susceptibility tests of irradiated 310S and Alloy 800H at 625 C (CA, EU) Modelling of fuel cladding degradation mechanisms (CA, EU) Round robin on general corrosion / stress corrosion cracking susceptibility tests to further assess facility-dependent effects (CA, CN, EU) Radiolysis and water chemistry (CA) Test on corrosion in non-isothermic loops: with an in-pile supercritical water loop (EU), supported by modelling (CA) out-of-pile loop tests (CN) Slide 13

14 Fuel Qualification Testing Project (Provisional) Scope In-pile fuel testing in a supercritical water loop with a capsule immersed into nuclear reactor Objectives Test of the licensing procedure, identify general problems Qualification of fuel rod and spacer manufacturing processes Test of measurement systems for supercritical water Test of fuel-cladding interaction Validation of material performance Validation of stress and deformation predictions Development of water chemistry strategy Validation of thermal-hydraulic predictions Validation of transient system code predictions Slide 14

15 FQT Project: Complete Facility Design at CVR Rez, Czech Republic Research reactor LVR-15 Test fuel assembly Dimensions Rod diameter Cladding thickness Rod pitch Wire thickness Wire pitch 8 mm 0.5 mm 9.44 mm 1.44 mm 200 mm Auxiliary systems of the supercritical water loop UO 2 enrichment Fissile power Linear heat rate 19.75% 63.6 kw 39 kw/m Slide 15

16 FQT Project: Heat Transfer Tests in out-of-pile loop at SJTU, China SWAMUP Supercritical Water Loop Steady-state experiments to measure the wall temperature for CFD validation Depressurization transient experiments to validate the system code at SJTU, China Slide 16

17 Contributions to SCWR Projects Slide 17

18 7th Int. Symposium on SCWR (ISSCWR-7) Successfully held on March 15-18, 2015, in Helsinki Hosted by VTT Technical Research Centre of Finland in cooperation with» Finnish Network for Generation Four Nuclear Energy Systems (GEN4FIN),» Generation IV International Forum (GIF),» International Atomic Energy Agency (IAEA)» Canadian Nuclear Society (CNS). 92 presentations were given by about 90 participants from 14 different countries 30 papers selected for publication in Journal of Nuclear Engineering and Radiation Science Slide 18

19 11 th GIF SCWR Information Exchange Meeting GIF SCWR SSC holds Information Exchange Meeting bi-annually among partners Provide in-depth descriptions in development and results on key technology areas IAEA has been invited to participate in the meeting Members of the coordinated research project participated The most recent meeting was held at NPIC in Chengdu, China Close to 80 participants from Canada, China, EU and Russian Federation Two days of presentations covering concept development and technology areas A technical tour of NPIC facilities Slide 19

20 Extension of the SCWR System Arrangement SCWR System Research Plan is being updated with 3 main directions: Thermal-Hydraulics & Safety (existing project) Materials & Chemistry (existing project) System Integration & Assessment (provisional project) covering» Fuel Qualification Testing» Concept Development & Assessment» Physics SCWR System Arrangement to be expired at the end of Nov 2016 Extension process to be initiated (similar to that for the SFR System Arrangement expired on 15 Feb 2016). Slide 20

21 Summary Several SCWR concepts are being pursued GIF R&D collaboration enhances understanding and expands knowledge bases that support the concept development process Several joint projects have been established leveraging expertise and resources among partners in the GIF SCWR system New partners joining these projects would further enhance the R&D capability and expedite the development process Slide 21

22 Super-Critical Water-cooled Reactor THANK YOU FOR YOUR ATTENTION! Slide 22