Tritium Control and Safety

Transcription

1 Tritium Control and Safety Brad Merrill Fusion Safety Program 1 st APEX Electronic Meeting, February 6, 2003

2 Presentation Outline Temperature & tritium control approach TMAP model of solid wall AFS/Flibe concept Tritium inventory MELCOR model of primary heat transport loop LOCA & LOFA temperatures Mobilization of activation inventories Dose results Initial CLiFF results Status of safety article

3 Temperature & Tritium Control Approach Thermal blanket of helium for both the primary and secondary cooling systems maintains internal temperatures at desired levels (> 500 C) and outer wall temperatures cool (~90 C) Tritium control & recovery by helium cleanup systems, cool outer walls, and aluminum pipes in Brayton cycle coolers

4 Primary Temperature Envelope Heat exchanger 100 C Triple wall pipe 500 C 700 C 1.4 m 1.0 m 0.7 m Secondary

5 Pebble Bed Modular Reactor Generator Pressure boundary (90 C) Turbocompressor Power turbine Helium Brayton cycle Inter-cooler Pre-cooler Recuperator

6 TMAP Model of Primary & Secondary Heat Transport Loops Secondary blanket High temperature shields Heat exchanger Secondary & pressure boundary walls Primary blanket Triple walled pipe Intercoolers Mixer Helium Cleanup

7 Tritium Inventory & Permeation Tritium Inventory AFS primary loop ~ 82 g, with 62 g in blankets. When inventory divided by AFS volume g/m 3 HYLIFE II ~ 1.4 g/ m 3 Flibe & helium ~ 1.1 g and 5.5 g respectively Limit 10 msv ~ 130 g stacked release, 15 g ground release Permeation Airborne Mixer, pipe walls, secondary pressure boundary ~ 1.2 g/y Limit for routine ground level release 1.3 g/y Water Brayton cycle coolers ~ 4.0 Ci/y; at coolant flow rate 13 pci/l per pass; 0.2 Ci/m 3 at end-of-life Limit 20,000 pci/l CANDU experience Ci/m 3 enter processing

8 Model Description The MELCOR model includes primary heat transport system (vacuum vessel -VV, blankets, high temperature shields, mixer, pipes, heat exchanger, accumulator, pumps, valves) The nuclear and decay heat values used in the model are the generated by Mohamed Sawan. The secondary side heat exchanger operates on a Brayton cycle VV water cooling during accidents by natural convection based on ITER-FEAT VV system - 80 kg/s MW for a 5 C water rise

9 Idaho National SCHEMATIC Engineering and Environmental OF Laboratory MELCOR MODEL Top View Based on quadrant of reactor IB VV IB HTS fl240 IB BKT fl210 fl245 Radial View OB BKT I fl260 OB BKT II fl270 OB HTS fl275 fl200 fl280 fl285 fl290 fl205 fl215 fl220 fl250 fl300 fl230 CV020 CV530 hs21400 CV700 hs21500 CV630 fl400 fl410 fl415 fl600 fl620 fl420 fl625 fl720 fl615 fl610 CV620 CV685 Drain Tank fl710 CV710 HTXS Secondary CV480 hs00690 CV470 hs00680 CV570 CV580 hs00780 CV036 fl265 fl295 hs00340 CV300 hs00341 CV310 hs00342 hs00324 CV321 hs00323 CV322 hs00322 hs00321 CV320 hs00320 hs00400 CV400 hs00401 hs00402 CV402 hs00403 CV401 hs00404 hs00430 CV440 hs00440 CV430 hs00450 CV450 hs00451 CV460 hs00452 CV521 CV520 fl225 CV540 hs00670 CV500 hs00650 CV600 CV640 hs00770 cv036 OB VV fl621 hs00350 hs00460 Mixer Triple wall pipe

10 LOCA Without VV Cooling Temperature (C) IB HTS IB VV OB FW BKT II OB HTS IB FW OB VV Time (d) Unacceptable AFS FW oxidation during bypass accident will occur at these temperatures

11 First Wall Temperature during LOCA and LOFA Temperature (C) LOCA without VV cooling LOCA VV cooling LOFA without VV cooling LOFA with VV cooling Time (d) Flibe provides thermal inertia during LOFA & VV natural convection appears to be able to remove decay heat (~3.6 MW max load)

12 Safety Assessment for Worst Case Bypass MELCOR LOVA model (VV, bypass duct, adjoining room, ventilation ducts) with FW temperatures from LOCA with VV cooling Radioactive inventories AFS specific dose - (varies with maximum of 9.1 msv/kg) (Mn %, Ca-45 15%, and Ti-45 14% of dose) Tritium specific dose - 67 msv/kg Flibe specific dose msv/kg Mobilization AFS from HT9 oxidation data for LOCA FW temperatures Permeation into vacuum vessel by transient TMAP analysis Flibe by evaporation in vacuum ( t < 100 s), and by diffusion through boundary layer (t > 100s)

13 Safety Assessment Dose at Site Boundary from Bypass Accident 1000 Mass released to environment 1.0 Site boundary dose Flibe 0.8 Tritium Grams released 10 AFS Tritium Dose (msv) AFS Time (d) Time (d) Total dose is 1.6 msv if release is stacked which is below 10 msv limit for no-evacuation plan 0.2 Flibe

14 Temperature & Tritium Control Approach For CLiFF/AFS Flinabe Design Primary piping (single wall) and heat exchanger are electrically heated and insulated to maintain temperatures at desired levels (> 300 C) and outer wall temperatures cool (~90 C) Internal vessel component heated by VV bakeout? Tritium control & recovery by intermediate salt loop between primary Flinabe loop and steam power cycle, pipe walls to be clad with aluminum. MSRE used a mixture of NaF and NaFB 2

15 Status of Safety Assessment for CLiFF Design Tritium inventories calculated by a TMAP model of primary cooling system Production rate 584 g/d Permeation from CLiFF implantation 220 g/d AFS inventory ~ 110 g (FW, BKT walls, piping), pebble will produce about 0.37 g/d Inventory divided by AFS volume ~ 1.1 g/m 3 Routine release not yet determined Performed MELCOR analysis of LOCA and LOFA AFS, Flinabe, and tritium mobilization rates need to be determined Worst case bypass accident not yet performed

16 First Wall Temperature during LOCA and LOFA Temperature (C) LOFA without VV cooling LOCA with VV cooling LOFA VV cooling LOCA without VV cooling Time (d) Flinabe decay heat a problem during LOFA, might either require draining of primary or in-vessel natural convection loops

17 Summary of Journal Article Outline developed and co-authors notified Still completing dose assessment for CLiFF design Tritium section of paper completed On schedule for March 1st?