1 / 13

Safety Issues Related to Flibe/Ferritic Steel Blanket and Vacuum Vessel Placement

Safety Issues Related to Flibe/Ferritic Steel Blanket and Vacuum Vessel Placement. Brad Merrill Fusion Safety Program. ARIES Project Meeting, Wednesday, May 7 th , 2003. Presentation Outline. Safety requirements ITER Confinement strategies ITER Confinement bypass accident scenarios

parry
Télécharger la présentation

Safety Issues Related to Flibe/Ferritic Steel Blanket and Vacuum Vessel Placement

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. Safety Issues Related to Flibe/Ferritic Steel Blanket and Vacuum Vessel Placement Brad Merrill Fusion Safety Program ARIES Project Meeting, Wednesday, May 7th, 2003

  2. Presentation Outline • Safety requirements • ITER Confinement strategies • ITER Confinement bypass accident scenarios • ARIES Compact Stellarator Bypass Accident Initiators • APEX radioactive inventories • APEX releases during a bypass accident and resulting site boundary doses • APEX waste disposal ratings • Summary

  3. Safety Requirements • The DOE Fusion Safety Standard enumerates the safety requirements for magnetic fusion facilities, two primary requirements • The need for an off-site evacuation plan shall be avoided, which translates into a dose limit of 10 mSv at the site boundary during worst-case accident scenarios (frequency < 10-6 per year) • Wastes, especially high-level radioactive wastes, shall be minimized, implying that all radioactive waste should meet Class C, or low level, radioactive waste burial requirements • To demonstrate that the no-evacuation requirement has been met, accidents that challenge the radiological confinement boundaries (e.g., confinement bypass accidents) must be examined.

  4. Cryostat Upper HTS vault Connecting ducts NBI cell Outboard baffle- limiter PHTS Rupture disks Divertor Suppression PHTS tank Basemat room Lower HTS vault 0 10 20 30 Schematic of ITER Confinement Barriers • Confinement of radioactive inventories by multiple barriers (defense in depth), primary boundary, secondary boundary • Vacuum vessel (VV) is part of primary confinement boundary Radius (m)

  5. Bypass accident initiators considered by ITER • Plasma Disruption • Disruption forces on VV fail a diagnostic duct that leads to a non-nuclear room, plus runaway electrons fail FW • Ex-vessel loss-of-cooling accident (LOCA) • Plasma continues to burn and FW fails by melting • Unmitigated Toroidal Field Coil Quench • Sensors fail to activate dump resistors • Quenched conductor melts by 20 s • Internal arcs form depositing ~10 MW per arc • 10’s of GJ of energy associated with magnetic field (44 GJ available) resistively dissipates in failed coil before arc leaves the magnet by way of magnet busbars • Arc travels along busbars and fails cryostat (hole > 2 m2) • Magnet melt is at high pressure(~120 bar) • Molten metal jet from the arcs create a hole in VV (~ 1 m2)

  6. Bypass accident initiators considered by ITER (cont.) • Unmitigated toroidal field coil quench (cont) (ITER FEAT calculations by N. Mitchell)

  7. ARIES Compact Stellarator Bypass Accident Initiators • For ITER, the plasma disruption initiated bypass scenario produced the largest off-site dose • ARIES-CS could potentially experience all three scenarios, with plasma disruption replaced by a rapid plasma bootstrap current quench possibly caused by FW failure and Flibe injection into plasma • If field coils are placed inside of the VV, then the unmitigated quench bypass accident will probably become the more severe accident because of multiple barrier failures • Magnet arcing/molten melt could fail blankets producing an in-vessel Flibe LOCA • Arc traveling along busbars could fail VV where busbars penetrate VV, releasing VV cooling water into plasma chamber and into cryostat • Water/Flibe interactions could result in steam vapor explosions and the mobilization of Flibe activation products • Busbar arc will eventually fail cryostat leading to a pathway for VV inventories to be released into the magnet power supply room (a non-nuclear room)

  8. APEX AFS/Flibe Blanket Radioactive Inventories • Inventories of concern • Advanced ferritic steel (AFS) activation products • Specific dose varies with time, maximum of 10.6 mSv/kg with Mn-54 at 26 %, Ca-45 at 15%, and at Ti-45 14% • Tritium • HTO specific dose is 77 mSv/kg • Flibe activation products • Specific dose - 0.32 mSv/kg with 99% F-18 • Mechanisms that can mobilize these inventories • AFS activation products by oxidation • FW high temperatures in air or water environment • Tritium by permeation into vacuum vessel • Evaporation of Flibe after a LOCA

  9. APEX Tritium Inventory & Permeation Issues • Problem is that tritium solubility & diffusivity are low in Flibe and high in AFS • Tritium Inventory • AFS primary loop ~ 82 g, with 62 g in blankets. • Flibe & helium ~ 1.1 g and 5.5 g, respectively • Neutron reactions with beryllium multiplier produces up to 2.1 kg over blanket lifetime • Tritium control & recovery • Helium purification systems • Cool pipe and pressure boundary walls • Aluminum pipes in Brayton cycle coolers • Beryllium tritium inventory reduced by bake-out, however tritium release temperature is initially ~850 C but will decrease to ~700 C after a fluence of 1.0 x1026 n/m2

  10. 1200.0 1000.0 LOCA without VV cooling LOFA without VV cooling LOCA VV cooling 800.0 Temperature (C) LOFA with VV cooling 600.0 400.0 0.0 5.0 10.0 15.0 Time (d) MELCOR FW Temperature during LOCA and LOFA Flibe provides thermal inertia during LOFA and VV natural convection appears to be able to remove decay heat (~3.6 MW max load), and low FW temperatures reduce oxidation

  11. Mass released to environment Site boundary dose 0 10 0.5 Flibe 0.4 -1 Tritium 10 0.3 Flibe -2 10 Dose (mSv) Mass released (kg) AFS 0.2 -3 10 Tritium 0.1 AFS -4 10 0 1 2 3 4 5 6 7 0.0 0 1 2 3 4 5 6 7 Time (d) Time (d) Dose at Site Boundary from Bypass Accident • Total dose after one week is 0.93 mSv (< 10 mSv no-evacuation plan limit) if release is stacked, must isolate within one week for a ground release • If Be bakeouts are successful, the blanket tritium inventory is 660 g. When this tritium is included with AFS inventories, the dose exceeds 10 mSv in six days for a stacked release, two days for a ground release

  12. APEX Waste Disposal Ratings • AFS can meet Class C limit • AFS structure WDR is 0.33-1.97 with Fetter limits, dominated by Tc-99 produced from Mo; reduce Mo content from 0.02% to <0.01% • Flibe WDR is 0.042 with 10CFR61 limits, major contributor is C-14 from neutron reactions with F

  13. Summary • Placing the field coils inside of the VV could lead to a severe bypass accident • APEX worst case bypass accident analysis shows that this low vapor pressure molten salt/low oxidation ferritic steel design has many safety advantages • Dose at site boundary is only 0.93 mSv after one week (< 10 mSv limit) for stacked release, facility must be isolated by one week for ground release • Ample time to manually operate plant remediation and isolation systems • Blanket and coolant will likely meet low level waste burial criterion

More Related