Preliminary Analysis of Loss of Vacuum Events in ARIES-RS David Petti and Brad Merrill Fusion Safety Program

The risk dominant events based on ITER experience are those associated with bypass of radiological confinement barriers Loss of Vacuum (reporting on this today) Ex-vessel Loss of Coolant without Plasma Shutdown In-vessel Loss of Coolant with Bypass

Loss of Vacuum Event Scenario Failure of double confinement in a penetration (perhaps heating and current drive system line) Air enters the plasma chamber Air extinguishes the plasma with a disruption that would mobilize dust and any “easily mobilizable” tritium Air exchanges between the plasma chamber and the “generic bypass” room that is communicating with the chamber because of density gradient between hot air inside and cooler air outside (stratified flow in the duct) Releases depend on duct size, orientation of duct, and nature of ventilation and filtration in the generic bypass room

Assumptions Leakage One air exchange per day. One volume loss per day at 400 Pa overpressure HVAC closed after 1 hour Assumptions In-vessel component surface areas exposed to free volume BKT 330 m2 DV 135.4 m2 Reflector 230 m2 LT shield 230 m2 VV 846 m2 Vacuum Vessel Free Volume 785 m3 Tritium Used 2.6 kg HTO Actual value TBD 0.02 m2, 10 m long In-vessel components Temperatures set to outlet design values Reduced linearly to inlet temperatures and then cooled at 30°C/hr until 225°C reached Tungsten Dust 100 kg at 2.11 microns, 10 kg at 0.1 microns - ITER spec GBR 5625 m3 2355 m2 wall area and 1500 m2 ceiling and floor

Other cases were studied to determine sensitivities 10 kg of dust/ 0.1 microns - used in ARIES-ST HVAC increased to exchange rate of 24 volume exchanges per day (non-nuclear type room) Generic bypass room (GBR) is elevated, raised from reactor mid-plane to that of “Other Function Areas” in ARIES-RS design. This elevated room (EGBR) is 4688 m3 with wall area of 750 m2 and ceiling plus floor area of 1250 m2 EGBR case with no isolation at one hour Base case with a larger duct break of 0.1 m2

MELCOR was used to analyze this event MELCOR is a fully integrated, engineering-level computer code that models the progression of accidents in light water reactor (LWR) nuclear power plants, including reactor cooling system and containment fluid flow, heat transfer, and aerosol transport. (Developed by Sandia National Laboratory)

Thermal Hydraulic Results

Radiological Results

Results compare favorably to no-evacuation release limits for tungsten

Summary of Findings Tungsten and tritium release results for 0.02 m2 break are fairly insensitive to room geometry and leakage/isolation dynamics. All releases are fairly low. Most of tungsten dust is in the VV, the duct and the floor of the GBR. The large density of tungsten promotes gravitational settling of the aerosol For the worst case 0.1 m2 breach, the release of tungsten is only 67% of the limit Tritium releases from the plasma chamber even in the worst case 0.1 m2 breach should be tolerable as long as the in-vessel mobilizable inventory remains below 3000 g HTO (~450 g-T).

Future Work Evaluate ARIES-AT relative to a LOVA (probably will be similar to ARIES-RS unless configuration around the tokamak changes drastically). Will have to deal with polonium issue. Evaluate ARIES-RS against the following two events (Li fire analysis will be required) Ex-vessel Loss of Coolant without Plasma Shutdown In-vessel Loss of Coolant with Bypass Qualitatively evaluate ARIES-AT for the Ex-vessel Loss of Coolant without Plasma Shutdown and In-vessel Loss of Coolant with Bypass events. Because of low chemical reactivity of LiPb, these events for ARIES-AT are probably not all that different from a LOVA