1 / 38

Japanese Perspectives on Liquid Blanket Research and Relating Collaboration

Japanese Perspectives on Liquid Blanket Research and Relating Collaboration. T. Muroga Fusion Engineering Research Center, National Institute for Fusion Science, Japan. APEX/TBM Project Meeting November 3-5, 2003, UCLA. Japanese Fusion Research Organizations.

viveka
Download Presentation

Japanese Perspectives on Liquid Blanket Research and Relating Collaboration

An Image/Link below is provided (as is) to download presentation Download Policy: Content on the Website is provided to you AS IS for your information and personal use and may not be sold / licensed / shared on other websites without getting consent from its author. Content is provided to you AS IS for your information and personal use only. Download presentation by click this link. While downloading, if for some reason you are not able to download a presentation, the publisher may have deleted the file from their server. During download, if you can't get a presentation, the file might be deleted by the publisher.

E N D

Presentation Transcript

  1. Japanese Perspectives on Liquid Blanket Research and Relating Collaboration T. Muroga Fusion Engineering Research Center, National Institute for Fusion Science, Japan APEX/TBM Project Meeting November 3-5, 2003, UCLA

  2. Japanese Fusion Research Organizations Two organizations carry out fusion research in Japan JAERI Mission given by the government Project-oriented ITER official contractor (at present) Universities (group of independent Professors) Mission defined by their own (more interest-oriented) Project by mutual agreement Scientific approach Playing complementary roles but sometimes causing problems (especially in making national decisions)

  3. Introduction of NIFS-FERS NIFS (National Institute for Fusion Science) is the inter-university research institute Coordination and enhancement of University research LHD as the core project Fusion Engineering Research Center in NIFS Established in 1999 Coordination and enhancement of University activity on Structural Materials Blanket (started 2001, activity still limited) SC system (nuclear technology related, 2003~)

  4. Introduction of NIFS-FERS (cont.) Present activities of NIFS-FERS Development of vanadium alloys (JAERI is the core for RAFS) Fabrication of reference ingots and characterization by universities MHD coating (started in 2000) Fabrication and corrosion tests Netronics (started in 2002) Liquid blanket, IFMIF IFMIF-Key Element Technology Verification Collaboration with universities (Li free surface in Osaka -) ITER-TBM needs coordination of Universities and thus potential major activity of FERC in the future

  5. Outline of Presentation What is agreed in Japan as strategy for liquid blanket research? Roadmap to powerplant Responsibility sharing between JAERI and Universities/NIFS Research emphasis in Universities/NIFS ITER participation/contribution (Including TBM) International collaboration

  6. General Classification into Two Lines for Fusion Development Fast realization of power demonstration Considered to be important for public support Based on modest progress in science/technology Relatively large budget allocation for near-term Project-oriented approach Currently JAERI is the core for this line Exploration of advanced system Long-term research including fundamentals Increasing attractiveness (cost, safety, environment) is thought to be crucial for fusion development Science-oriented approach Currently University/NIFS is the core for this approach

  7. EU Strategy Has Also Two Lines Reference system Advanced system Lackner, ICFRM-10

  8. Two Lines for Materials/Blanket in Japan Fast realization line RAFM/water as a reference system RAFM/Supercritical water (optional 1) ODS/Supercritical water (optional 2) Relatively large budget allocation for the development RAFM test planed to be dominant in early stage of IFMIF Advanced line Presently liquid blanket systems (V/Li and Flibe) and SiC/He Focusing on fundamentals and key feasibility issues Science-oriented approach University/NIFS is the core for this approach Major subjects of JUPITER-II

  9. Two Lines for ITER-TBM(preliminary discussion) Fast realization line RAFM/water TBM Efforts focused on Day One TBM Contribute to licensing First Power Generation Plant together with early IFMIF data Advanced line Liquid blanket systems (V/Li and Flibe) and SiC/He Plan to start TBM either from Day One or in the later phase of ITER Agreed to keep these activities irrespective of the selection on Day One TBM

  10. Approximate calendar year 2015 2020 2030 2040 Advanced Powerplant Design Power Generation Plant Design Construction Operation (Licencing) (Blanket test) ITER Blanket Module Test Materials and Blanket System Development Reference Material (RAFM) and System Advanced Materials (V-alloy, SiC/SiC --) and System IFMIF Irradiation Test, Materials Qualification and System Performance Test (Staged construction and operation) Schematic Roadmap for Materials and Blanket Development in Japan Fast realization line (Currently JAERI leadership) Advanced line (Currently NIFS/University leadership)

  11. From Japan, only solid breeders were proposed to ITER via JAERI Participation to scientific aspects of ITER research by NIFS/ Universities is being enhanced A NIFS-collaboration activity started in 2002, in which liquid blanket test module is explored Recent Activity of ITER-TBM in Universities/NIFS 1995 2001

  12. Examination of Li/V first and then followed by Flibe Support from JAERI First output expected in 2004 NIFS Collaboration Activity for ITER-TBM

  13. Feasibility of no-Be and natural Li blanket Use of 7Li reaction for enhancing TBR in contrast to Russian Be+6Li enriched TBM Validation of neutronics prediction Technology integration for V-alloy, Li and T Purpose of Li/V ITER-TBM(current discussion)

  14. ITER with Li/V self-cooled blanket - MCNP calculation by T. Tanaka (NIFS) - [ Inboard ] 40 cm SS, H2O A Vacuum vessel Blanket FW Plasma B SS (60%), Li coolant (40%) V-4Cr-4Ti walls, Natural Li [ Outboard ] Coil structure 40 cm 1 m Vacuum vessel + Filler SS, H2O SS, H2O Center solenoid A Blanket Vacuum vessel FW Blanket A : Standard ITEF-FEAT blanket B B : ITER with V/Li full blanket Input geometry for MCNP calculation * V-4Cr-4Ti walls, Natural Li SS (60%), Li coolan (40%) (*Dimensions from ITER Nuclear Analysis Report)

  15. ITER with Li/V self-cooled blanket - Local TBR - Local TBR (Full Coverage)* (* JENDL 3.2) (a) Inboard (b) Outboard FW FW Blanket Blanket Filler Filler Distribution of tritium production rate ■ Significant contribution of 7Li to TBR

  16. Neutron spectrum at first wall of Standard and V/Li Blanket Comparison of Neutron Flux at Outboard First Wall Cross Section for Tritium Production (JENDL 3.2) ■ Significant difference between thermal neutron component in ITER-FEAT and ITER-Li/V ■ Thermal neutron should be shielded in the TBM area of ITER-FEAT for the purpose of simulating V/Li blanket condition

  17. Russian Li/V self-cooled test blanket module- Structure - 505 WC Shield (Reflector) Be multiplier SS(60%) + H2O(40%) Plasma 1720 V-5Cr-5Ti Li layer (6Li : 90%) Structure of Russian Li/V TBM (Unit : mm) ■6Li enriched coolant(7.5 % ==> 90%) ■ Li layer x 2, Be multiplier ==> 6Li (n, a) T ■ Maximize the 6Li reaction to demonstrate DEMO reactor breeding tritium by 6Li

  18. Be WC Plasma SS+H2O Li layer (1) Li layer (2) [6Li : 90%] Russian Li/V self-cooled test blanket module- Tritium production - Total : 0.09 (g/FPD) SUS + H2O Plasma Li layer (1) Li layer (2) TBM surface Tritium production rate in Li layers and contribution of 6Li and 7Li SS316 TBM frame Li/V TBM

  19. Tentative design of Li/V self-cooled TBM by NIFS/Universities 505 Verification of (1) Coolant circulation (2) MHD coating SS(60%), H2O(40%) Plasma Li layer V-4Cr-4Ti Verification of (1) Neutron transport (2) Tritium production from 7Li SS316 TBM frame 210 1720 210 470 Plasma Inlet/outlet pipes SS(60%), H2O(40%) Li : ~0.027 m3 Li/V TBM (Unit : mm) Tentative design of Li/V TBM ■ Verification of TPR for 7Li ■ Thick Li tanks for verification of neutron transport

  20. Tentative design of Li/V self-cooled TBM - Tritium production - Covering by B4C ■ For verification of tritium production from 7Li (n, na)T reaction Plasma SS(60%), H2O(40%) - Reduction of thermal neutrons by B4C shielding Li layer (1) (2) (3) (4) (5) Li layer (1) Li layer (1) (2) (2) (3) (4) (5) (3) (4) (5) Contribution of 7Li to tritium production Tritium production rate in Li layers

  21. Experimental parameter for Li/V TBM - Adjustment by B4C shield - 10 cm in front side Li/V TBM 10 cm in rear side 10 cm in front side Li/V blanket 10 cm in rear side Russian TBM Changes in contribution of 7Li by B4C covering ■ Contribution of 7Li to TPR can be adjusted by thickness of B4C shield

  22. Future Participation to ITER-TBWG(discussion not started) Discussion on participation of University/NIFS to ITER-TBWG will start soon Possible options may be Tune the present blanket activity to TBWG schedule Start engineering design for V/Li TBM Concept definition and start engineering design for Flibe TBM Keep the present pace with weaker interaction with TBWG In this case, we will not strongly propose Day One TBM Keep the present advance blanket research activities irrespective of the selection on Day One TBM

  23. JUPITER-II JUPITER-II is a mission-defined collaboration program Advanced blanket (in contrast to JAERI’s FS/water) Task plan and Check and Review Use of core facilities (HFIR, STAR --), which are unavailable in Japan, is the rationale for the collaboration (transferring budget from J to the US) Major change of the framework need re-evaluation by standing committees and will face a risk Most Japanese JUPITER-II participants have strong scientific interests in the present tasks and have small incentive to make extensive change in the framework

  24. JUPITER-II Possible Fine Tuning(unofficial, Muroga private idea) “We cannot propose any concept for ITER-TBM at present with the lack of corrosion data” (Sze-Muroga e-mail agreement) Shift some effort from vanadium irradiation to MHD coating/corrosion MHD coating is the critical issue for both long term blanket development and entry to Day One TBM REDOX,Flibe-materials interaction should be enhanced MHD related design activity should be enhanced Lenient requirement to MHD coating for V/Li However, HFIR and Tritium activity must be maintained because of program need and participants incentives

  25. Comment/questions to US Discussion on TBM Selection What is the philosophy of selecting TBM? Technical feasibility and ? Japan : Roadmap to Powerplant What is the community selecting TBM? Liquid breeder for Japan : University/NIFS including B, M, T, S – (If materials people are not involved heavily, the impact of the decision on materials program must be small) Why two options (only because of budget?) Number of available port no longer the factor What is the fate of the concepts not selected for the first-day TBM? (Longer-term strategy)

  26. End of presentation

  27. MHD Coating – Necessity– Magnetic Field Duct MHD Pressure Drop ・Load to pumping system ・Force to structures Li Flow Force Pressure Drop： proportional to Flow length、Velocity、B2、Duct thickness、Conductivity of Li and Duct Insulator coating inside the ducts a possible solution

  28. Ｖ２Ｏ５ Ｌｉ２Ｏ 100℃ 1 x 10-3 2 x 10-3 3 x 10-3 １／Ｔ (K) MHD Coating Candidates (1)–Free Energy Stable ceramics in a quite reducing condition Selection from the free energy data CaO、Y2O3、Er2O3、 CaZr(Sc)O3、 AlN、BN

  29. Potential candidates Y2O3 Er2O3 AlN with N control CaZr(Sc)O3 (~700C) others 10mm/y MHD Coating Candidates (2)– Bulk Compatibility Japan-US JUPITER-II Collaboration (Pint, Suzuki et al. 2002)

  30. Development of coating technology RF-sputtering EB-PVD Arc Plasma Deposition Characterization of the coating Resistivity High temperature stability Compatibility with Li Radiation induced conductivity In-situ coating technology MHD Coating DevelopmentPresent Efforts Japan-US JUPITER-II Collaboration (Suzuki, Pint et al. 2003）

  31. The in-situ coating method has advantages as, possibility of coating on the complex surfaceafter fabrication of component potentiality to heal the cracks without disassembling the component CaO coating has been explored In-situ Coating

  32. It was found that the CaO coating, after formation, dissolved at high temperature (600, 700C) CaO bulk is inherently not stable in pure Li at high temperature, continuous supply of oxygen is necessary to maintain the coating Er2O3 is much more stable at high temperature It is expected Er2O3, once formed, be stable in Li for a long time Er2O3 is stable in air, combination of dry-coating and in-situ coating is more feasible CaO Er2O3 Problems of the CaO Coating and New Effort on Er2O3

  33. Er2O3 layer was formed on V-4Cr-4Ti by oxidation, anneal and exposure to Li (Er) at 600C The coating was stable to 300 hrs The resistivity was ~1013 ohm-cm 5 5 x 10 Er2O3 - layer - 0030_1.PRO x 10 Er2O3 - layer - 0035_1.PRO 2 2 O1s O1s Er4d 1.5 1.5 Er4d V2p3 V2p3 Intensity Intensity 1 1 0.5 0.5 0 0 0 5 10 15 20 25 30 0 5 10 15 20 25 30 Sputter Time (min) Sputter Time (min) 5 5 x 10 Er2O3 - layer - 0067_1.PRO x 10 Er2O3 - layer - 0062_1.PRO 2 2 O1s O1s Er4d Er4d 1.5 1.5 V2p3 V2p3 Intensity Intensity 1 1 0.5 0.5 0 0 0 5 10 15 20 25 30 0 5 10 15 20 25 30 Sputter Time (min) Sputter Time (min) In-situ Er2O3 Coating on V-4Cr-4Ti Oxidation and anneal at 700C for 16 hr Oxidation only Oxidation at 700C ~100 nm 6 hr 1 hr Yao. 2003 XPS depth profile after exposure to Li (Er) at 600C for 100 hr

  34. Requirement to the coating performance depends strongly on the design System design is necessary to quantify the requirement to the coating Clever design would make the requirement lenient New idea of coating will be obtained by collaboration with design Laminar coating structure, etc. Need for Collaboration with Design People

  35. For a single pipe, with a perfect insulating coating, the allowable crack fraction was <10(-7) For a real coating, 10(-2) is achievable, while 10(-4) might be achievable.  If we start with a poor coating, the allowable fraction can be higher, maybe 10(-4), with a higher MHD pressure drop. There are other ways to increase the allowable crack fraction, such as change the aspect ratio of the channel, change the boundary conditions of the flow channel. The boundary condition of the flow channel, such as the contact resistance between the fluid and the wall, may have major impact on the crack fraction.    The change of the designs may have major impact on the crack allowance. Meeting Summary for Crack Fraction Allowance for the MHD Coating (Sze Aug.03)

  36. Experimental examination of the resistance between the (flowing) Li and the wall covered with cracked coatings at high temperature is of high priority.  The goal of the in-situ healing may be set to increase the resistivity of cracked area from complete conduction by 4 order of magnitude Increased collaboration between materials and design people in Japan Impact of Sze Summary on the Coating Development in Japan

  37. Optimization of channel structure for reducing the requirement to the coating Coating may be necessary only on limited flat surfaces Insulator ribs may be inserted instead of coated ribs/walls Other suggestions on laminar coating structure, enhanced heat transfer, etc Design Effort to Reduce Requirement to the Coating (Hashizume)

  38. Collaboration by Materials, Blanket and Design people are increasing on V/Li system in Japan Progress in developing vanadium alloys toward engineering maturity Enhanced Li technology by IFMIF-KEP Increased accessibility for the liquid blanket people to ITER-TBR Collaboration on MHD-coating development by materials and design people One of the goals of the collaboration is to propose V/Li ITER-TBM The collaboration is enhancing research for other advanced blanket systems (Flibe --) The collaboration covering Material, Blanket and Design people in the US will accelerate the progress, and should be enhanced in the framework of JUPITER-II Summary

More Related