Overview of the INL Fusion Safety Program Pb

Overview of the INL Fusion Safety Program Pb. Li-T Workshop 11 -12 June 2007 Idaho Falls, ID

Fusion Safety Program Outline • • STAR Dust Tritium Material Interactions and Permeation Fusion Safety Codes Risk Waste Management Safety Design Analysis Summary Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 1

Fusion Safety Program Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 2

Fusion Safety Program STAR Floorplan Layout Glovebox TCS Tritium SAS Stack monitor Chemical reactivity D-ion implantation TPE 2 Li. F-Be. F 2 preparation, purification and testing MS tritium exp MS corrosion exp Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 15, 000 Ci tritium limit Segregation of operations Gloveboxes and hoods Tritium cleanup system Once-through room ventilation Slide 3

Fusion Safety Program Star Tritium Storage and Assay System Secondary inlet Primary outlet U-bed-1 U-bed-2 SAS glovebox Setup Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 SAS manifold with U-beds SAS manifold and vacuum pumps Slide 4

Fusion Safety Program STAR Tritium Cleanup System (TCS) Outlet IC Inlet sample loop Condenser Control Heat exchanger MS water Mole sieve bed Blower Catalyst beds Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 5

Fusion Safety Program Tritium at STAR • Useable tritium inventory now 1300 Ci – – – 300 Ci in equimolar H 2: D 2: T 2 calibration standard – next shipment (1000 Ci) expected in Fall 2007 1000 Ci T 2 available for experiments shipments from SRS limited to 1000 Ci with standard TYPE-A shipping container Shipping Vessel (1 available) Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 6

Fusion Safety Program May 1995 - First tritium plasma experiments at TSTA September 2000 - Final tritium plasma experiments at TSTA History of TPE March 2002 - Extraction, loading, and transport April 8, 2002 - Depart LANL � February 2001 - Begin D&D efforts of TPE December 2001 - Final pump out of system; close all valves January 2002 - Preparation for shipment Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 April 10, 2002 - Arrival at INL STAR Facility Summer/ Fall 2002 Uncrate and decon ancilliary components January 2003 Modify plans for location of experiment, decide on Perma. Con structure, initiate facility design changes. June 2003 - Perma. Con installed, TPE glovebox uncrated. 2004 - Reassembly and system interface design activities Spring & Summer 2005 Electrical Service re-design & construction, experiment & facility interface assemblies completed Fall 2005 - Integrated systems testing initiated December 2005 - First plasma testing (non-tritium) Slide 7

Fusion Safety Program Planned Research Agenda for TPE • Study uptake, retention and permeation in PFCs – Measurement of bulk tritium transport properties (diffusivity, solubility, dissociation/recombinatino rates) – Monomaterials (Be, W, C) – Mixed materials – Bonded and/or duplex structures – Effects of neutron dose and irradiation temperature on tritium trapping • Certification of these structures for ITER • Use of tritium in the plasma will enable low level measurements needed for such research Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 8

Fusion Safety Program Flibe Tritium Experiment Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 • Tritium provided in pressurized vessel containing D 2/T 2 mixture • Glovebox setup to contain potential leaks • Localized tritium cleanup will be connected • GC column accurately measures H concentration; tritium release measured with ion chamber • well-characterized with D 2; now experiment with T 2 Slide 9

Fusion Safety Program Permeation Coating Barrier Experments Thermal Cycle Performance of He Pipe Permeation Barriers • simulates thermal stress degradation of permeation barrier coatings for He pipes • configuration matched to TBM design for coated components • utilize tritium for barrier technology qualification • external thermal cycles followed by testing in permeation rig for integrated effects • in-situ thermal cycling in permeation rig for barrier dynamic response Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 10

Fusion Safety Program Dust generation ITER Key Issues: chemical reactivity, radioactivity content, dust explosions. Science: understand model underlying formation mechanisms to estimate inventories expected in fusion ITER Dust Strategy from EDA Demonstration of the filtered vacuum collection technique Key scientific issues needing resolution Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 11

Fusion Safety Program Dust Characterization Particle Size Distribution, Specific Surface Area, Surface Mass Density, Composition, Shape and Tritium Content Specific Surface Area Comparison of Size Distributions CMD (µm) + GSD Machine Lower Regions Middle Regions Upper Regions 0. 89 + 2. 92 DIII-D 0. 66 + 2. 82 0. 60 + 2. 35 TFTR 0. 88 + 2. 63 1. 60 + 2. 33 Alcator-Cmod 1. 58 + 2. 80 1. 53 + 2. 80 JET 27 + (-) - - TEXTOR 5 -20 + (-) - - Tore Supra 2. 68 + 2. 89 2. 98 + 2. 94 3. 32 + 2. 94 ASDEX-Upgrade 2. 21 + 2. 93 3. 69 + 2. 81 3. 59 + 3. 08 LHD 8. 59 + 2. 67 6. 31 + 2. 39 8. 73 + 2. 09 NOVA 1. 12 + 1. 90 0. 76 + 2. 03 0. 90 + 1. 93 1. 22 + 2. 03 Dust Size versus Surface Mass Density R&D continues to resolve the issues Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 12

Fusion Safety Program Research Agenda for Dust • • • Continued characterization in existing tokamaks Mobilization testing Chemical reactivity of dust in grooves Dust explosion testing Monitoring and removal evaluation Improved strategy for ITER Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 13

Fusion Safety Program Major Accomplishments in Risk Assessment for Fusion • • Work on component failure rate data to support quantitative safety assessment continues to be very successful. – Initially, component failure rates were collected from handbooks and applied to fusion. – Now, through IEA Task 5, we collect fusion facility operating experience data from TLK, TPL, and the former TSTA; and tokamak data from DIII-D and JET. – Independent data sets validate the failure rate values. Occupational safety is a new area for risk assessment. – ITER IT has plans to perform a room-by-room overall assessment of the ITER facility to identify occupational hazards – Occupational injury rates have been collected from several US tokamaks and large particle accelerators – WE-FMEA method was developed to address highly hazardous equipment failures in a fusion facility, such as high energy pipe breaks that have caused worker fatalities in the power industry Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 14

Fusion Safety Program The Hazard Zone of the Worker Exposure Failure Modes and Effects (WE-FMEA) Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 15

Fusion Safety Program INL FSP Support of the ITER Project • The FSP is supporting the ITER Project through three Implementing Task Agreements (ITA): • Fusion Safety Code Support – – – • Magnet Safety – – – • Provide International Team (IT) with the latest fusion versions of the MELCOR and ATHENA codes, documentation, validation, and support and assistance at operation of the codes. Delivered MELCOR 1. 8. 5, upgraded ice layer model for cryogenic surfaces, and developed a beryllium dust layer oxidation model for MELCOR Assist the ITER IT in producing the QA documentation for MELCOR and safety analyses for ITER’s Report on Preliminary Safety (RPr. S) Update the MAGARC code to current ITER design for TF and PF coils, include new R&D results on insulation failure behavior at elevated temperatures, apply the MAGARC code to various ITER magnet safety studies Upgraded insulation and magnet parameters, implemented arc limit model, applied MAGARC to ITER-FEAT TF and PF magnet unmitigated quench events Develop magnet Busbar arc capability for MAGARC Dust Issues – – Characterize and quantify amounts generated during ITER operation Reactivity/explosivity, mobilization, and transport during accident conditions Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 16

Fusion Safety Program MELCOR Code Heat Structure Dust Layer Oxidation Model • • A beryllium dust layer oxidation model was developed for ITER to simulate oxidation of a dust layer of dust that has settled onto a heated surface inside of a fusion device This model is based on measured oxidation reaction rates for fully dense beryllium, binary gaseous diffusion of oxygen or steam into the dust layer, and BET measured specific surface area for beryllium dust Application is for slow vacuum vessel pressurization events from in-vessel loss-of-cooling accidents (LOCAs) and loss-of-vacuum accidents (LOVAs) 10 0 INL 92 88% dense Beryllium oxidation rate (kg/m 2 -s) • 10 MELCOR dust layer model ( Dust=0. 7 g/cm 3, dp= 20 mm) -3 INL 98 88% dense INL Dust 10 -6 10 -9 10 -12 INL fully dense 5 10 15 20 10, 000/T (K) Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 17

Fusion Safety Program MAGARC Poloidal Field Coil Development • The MAGARC code was recently modified to analyze unmitigated quench events in ITER poloidal field (PF) magnets • This modification include the electrical characteristics of the twoin-hand winding pair of the ITER PF coils and limits on the number of arcs that can form in the magnet during unmitigated quench events based on an energy minimization principle. Radial direction Winding pair Axi n al d ial d 500 1000 1500 irec tion Inline arcs voltage drops 0 Voltage (V) Rad Axial direction tio irec 0. 0 0 0. 2 5 He 0. 5 igh 0 Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 t (m ) Time 110 s 0. 7 5 0. 72 4 5 0. 1. 0 0 0. 36 8 m) 0. 1 th ( 0 d 0 i. 0 W Slide 18

Fusion Safety Program MAGARC code application to unmitigated quench events in ITER poloidal field (PF) magnets 3000 2000 1500 1000 0 150 300 Time (s) 450 6 4 2 0. 6 0. 4 0. 2 0 150 300 Time (s) 450 600 Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 0. 0 0 150 300 Time (s) 450 600 0. 12 Melt volume (m 3) Inline Radial Axial 2. 0 0. 0 600 0. 8 Quench fraction Number of arcs 8 3. 0 1. 0 500 0 0 4. 0 Coil current (k. A) • Lead voltage drop (V) MAGARC Poloidal Field Coil Application 0 150 300 Time (s) 450 600 0. 08 0. 04 0. 00 0 150 300 450 Time (s) 600 Slide 19

Fusion Safety Program Future Safety Code Activities • MAGARC capabilities will be expanded to treat arcs in magnet busbars by including the magnetic effects of the arc that forms between the leads of a busbar. As part of an international collaboration, this new capability will be validated against data that has recently been obtained from the MOVARC experiment at Fz. K in Germany • Work with the ITER IT to provide the necessary quality assurance documentation required for ITER licensing for the MELCOR code • Continue in support of the licensing process for the US DCLL TBM to complete Dossier on Safety for this TBM concept Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 20

Fusion Safety Program MELCOR Code Applied to ARIES-CS has Dual Cooled Liquid Lead Lithium (DCLL) Blanket • This blanket concept employees reduced activation ferritic steel (RAFS) for the structure, cooled by 8 MPa helium, and a self-cooled breeding zone cooled by flowing Pb. Li. Helium pressurization accidents will be a concern for this concept, with the reactor cryostat serving as secondary radioactivity and pressure confinement Because stellarator plasmas do not disrupt, the beyond design basis bypass accident (BDBA) of concern may be a heat exchanger tube breach in the Pb. Li/Brayton cycle system, leaking 10 MPa helium into the blanket causing a blanket break and pressurization of the vacuum vessel • • Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 0. 25 Pressure (MPa) • 0. 20 Vacuum Vessel Cryostat 0. 15 0. 10 0. 05 0. 00 3600 3700 3800 Time (s) 3900 4000 Initial results for in-vessel FW helium LOCA Slide 21

Fusion Safety Program MELCOR Code Applied to US Test Blanket Module Safety Assessment • • • Evaluate consequences to ITER from accidents in the proposed US DCLL Test Blanket module (TBM) To date three accident scenarios have been investigated: – In-vessel TBM coolant leaks – In-TBM breeding zone coolant leaks – Ex-vessel TBM cooling system leaks No significant impacts on ITER safety have been identified, but assessment is still ongoing Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Internal Pb. Li flow • All FS structures are He-cooled by 8 MPa • Pb. Li self-cooled flows in poloidal direction He out He in Pb. Li concentric inlet/outlet pipe Slide 22

Fusion Safety Program Recent Trends in Radwaste Management • Options: – Disposal in repositories: LLW (WDR < 1) or HLW (WDR > 1) – Recycling – reuse within nuclear facilities (dose < 3000 Sv/h) – Clearance – release slightly-radioactive materials to commercial market if CL < 1. • Tighter environmental controls and the political difficulty of building new repositories worldwide may force fusion designers to promote recycling and clearance, avoiding geological disposal No radwaste burden on future generation. • There’s growing international effort in support of this new trend. • Recycling may not be economically feasible for all fusion components. • Recycling of liquids and solids may generate limited amount of radioactive waste that needs special treatment. Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 23

Fusion Safety Program S&E considerations are critical for the success of IFE • IFE has both radiological and toxicological hazards: – Tritium fuel, activated structural material, activated dust, activated coolants or coolant impurities, and activated gases – Chemically toxic materials (i. e. : Hg, Pb) • Energy sources that can mobilize these hazardous materials include: – chemical energy, decay heat, pressure energy, electrical energy and radiation • In the US, current IFE S&E activities, are focused on a few programs: – The High Average Laser Program (HAPL) – The Z-IFE Program – The National Ignition Facility (NIF): not exactly an IFE program but closely linked to the future of IFE research Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 24

Fusion Safety Program Summary • US S&E research continues to help improve fusion facility design in terms of accident safety, worker safety, and waste disposal. • The R&D underway and currently planned in the areas of dust and tritium source terms will answer important questions for ITER and future machines. • Regulatory approval of ITER and the associated verification and validation activities for our fusion safety codes and risk and reliability methods will provide greater confidence in application of these tools to evaluate public and worker safety of future fusion facility designs. • The resurgence of nuclear fission reactor construction activities worldwide will cause increased attention to waste management issues associated with nuclear power which in turn should help fusion as it develops a long term waste management strategy consistent with on-going US regulation. • Safe and environmentally sound operation of both ITER and NIF will be important public demonstrations of the S&E potential of fusion. Pb. Li-T 2007, Idaho Falls, ID, 11 -12 June 2007 Slide 25