Safety Issues in TBM Program

Transcription

1 Idaho National Engineering and Environmental Laboratory Safety Issues in TBM Program Brad Merrill, Dave Petti 1 Hans-Werner Bartels 2 1 Fusion Safety Program 2 ITER IT, Safety Group APEX/TBM Meeting, UCLA, November 3-5, 2003

2 Idaho National Engineering and Environmental Laboratory Presentation Outline General Safety Framework for Test Blanket Modules (TBMs) Specific safety requirements for TBMs Safety results for already proposed ITER TBMs What these safety results mean to US Proposal Team (PT) TBMs What we learned from APEX and how this translates into US PT TBM selection process Conclusions

3 Idaho National Engineering and Environmental Laboratory General Safety Framework for TBMs Safety approach chosen by ITER is the development of Generic Safety Dossiers for all proposed TBMs The Contents for Safety Dossiers can be found in a MEMO located on ftp.itereu.de server * * Password protected server, contact H.-W. Bartels for additional information

4 Idaho National Engineering and Environmental Laboratory TBM:Contents of Safety Dossier General Demonstration of acceptable safety Consistent with ITER Preliminary Safety Analysis Report (PSAR) Either stand alone or integrated Note: in past safety documents TBMs were documented as stand alone Technical information for dossier provided by blanket proponent Common safety and design aspects from Plant Description Document (PDD, etc.) Detailed design information can be incorporated by reference These notes based on interpretation from site candidate inputs

5 Specific Idaho National Engineering safety and Environmental requirements Laboratory TBMs (1) (see PSR 2001,par of FDR-2001) ITER vacuum vessel (VV) and pressure suppression system are the the first confinement barrier for the TBMs All ex-vessel parts of cooling and other auxiliary systems are part of the first confinement barrier Test blanket cells shall provide the second confinement barrier Chemical reactions between coolant, air and breeder/multiplier material shall be limited so that the confinement function (i.e. of both barriers) is not threatened TBM shall be recessed 50 mm from the first wall of basic machine In-vessel liquid Li restricted to less than 35 liters to limit hydrogen production from H 2 O reactions to 2.5 kg (VV flammability limit 10 kg)

6 Idaho National Engineering and Environmental Laboratory Specific safety requirements TBMs (2) LiPb should be limited to 0.28 m 3 to limit hydrogen production to 2.5 kg from H 2 O reactions. Alternatively, detailed analysis of water/lipb interaction should be performed Special consideration for Li fires in local test module confinement shall be made Intermediate cooling loops are necessary for liquid Li system Beryllium of the first wall of a test module should be limited to 10 kg to limit hydrogen production to 2.5 kg from H 2 O reactions Decay heat removal should be achieved by thermal radiation to the basic machine

7 Topics Idaho National Engineering to be and addressed Environmental Laboratory are essentially all topics in Generic Site Safety Report Safety Approach Safety Design Radioactivity and Energy Source term Normal Operation Waste and Decommissioning Occupational Safety Analysis of Reference Accidents Ultimate Safety Margin External Events Sequence Analysis Safety models and codes

8 Idaho National Engineering and Environmental Laboratory Analysis of Reference Accidents In-vessel TBM coolant leak analysis to demonstrate: A small pressurization of first confinement barrier (i.e. ITER VV) Passive removal of TBM decay heat Limited chemical reactions and hydrogen formation Coolant leak into TBM breeder or multiplier zone analysis to assess: Module and tritium purge gas system pressurization Chemical reactions and hydrogen formation Subsequent in-vessel leakage Ex-vessel LOCA analysis to determine: Pressurization of TBM vault Behavior of TBM without active plasma shutdown

9 Accident analysis performed so far: Idaho National Engineering and Environmental Laboratory Summary of significant results for in-vessel TBM coolant leak European helium cooled pebble blanket (EU HCPB) Plasma shutdown by ingress of helium coolant (no disruption and as a result no ITER FW water coolant leak) VV pressurization by helium (inventory of only 15 kg) was small ~14 kpa Demonstrated adequate decay heat removal by thermal radiation to ITER in-vessel structures Passive Decay Heat removal for EU Proposal Team (PT) HCPB TBM ITER FW TBM FW

10 Summary Idaho National Engineering of significant and Environmental Laboratory results for invessel TBM coolant leak (cont.) Japanese water-cooled ceramic TBM Plasma shutdown by ingress of coolant VV pressurization by water (inventory of ~ 2000 kg) is ~70 kpa H 2 produced by chemical reactions with beryllium of ITER FW and TBM was acceptable Russian lithium self-cooled TBM 18 liters of lithium drains into VV port extension for TBM If plasma is not passively shutdown then TBM FW melt will occur (~1890 C for V-Cr-Ti alloy) adding an additional 8.6 liter of lithium into VV

11 Summary Idaho National Engineering of significant and Environmental Laboratory results for invessel TBM coolant leak (cont.) Russian lithium self-cooled TBM (cont.) Plasma disruption is assumed to follow TBM FW melt, which fails ITER FW and introduces water coolant into VV Lithium reacts with water to produce 2.5 kg of H 2 (within limits), a reaction zone temperature 1000 C, and heat up of VV in-vessel structures is ~ 110 C Because VV pressure suppression system will activate, the VV pressurization is not expected to be much higher than that for ITER FW loss-of-coolant accident (LOCA)

12 Summary Idaho National Engineering of significant and Environmental Laboratory results for invessel TBM coolant leak (cont.) European water-cooled lithium lead TBM Rupture of TBM FW is assumed which leads to passive plasma shutdown by water coolant and VV pressurization similar to Japanese ceramic TBM PbLi from TBM spill into VV is limited to 0.28 m 3, and the complete reaction of the PbLi is assumed to occur producing 2.5 kg of H 2 The mass of PbLi spill is limited by an isolation valve (active system) that separates the TBM inventory from that of ex-vessel components

13 Accident analysis performed so far: Idaho National Engineering and Environmental Laboratory Summary of significant results for In-TBM coolant leak Japanese water-cooled ceramic blanket Structural integrity of TBM box and tritium purge gas system maintained by use of passive pressure suppression system (peak ~ 0.46 MPa) H 2 from chemical reactions with beryllium remains within TBM system Mobilization of tritium and activation products (small) In-TBM LOCA for WCPB TBM of JA PT: Effect of Suppression Tank Size and Condensation

14 Summary Idaho National Engineering of significant and Environmental Laboratory results for In- TBM coolant leak (cont.) European helium-cooled ceramic breeder TBM Failure leads to pressurization of TBM and tritium extraction sub-system to 8 MPa within between 2 and 13 seconds depending on internal piping break size The design approaches to limit the consequence of this accident are Pressure relief of module via a rupture disk leading into ITER VV Fast acting isolation valve for the tritium extraction sub-system

15 Summary Idaho National Engineering of significant and Environmental Laboratory results for In- TBM coolant leak (cont.) European water-cooled lithium lead TBM Water chemically reacts with LiPb generating heat and hydrogen Determining the extent of this reaction is an ongoing EU R&D program For now, all Li in the LiPb is assumed to react but analysis of the event shows that the TBM box can handle the pressure and confine the products of the chemical reaction Russian lithium self-cooled TBM This type of accident is not applicable to a self-cooled design concept

16 Accident analysis performed so far: Idaho National Engineering and Environmental Laboratory Summary of significant results for loss of flow or ex-vessel coolant leak European helium-cooled ceramic breeder TBM Leads to release of helium into the large (~40,000 m 3 ) tokamak water cooling system (TWCS) vault resulting in an insignificant vault pressure increase If the active plasma shutdown system does not actuate due to high temperature TBM signal (a bounding event), then the TBM FW will likely fail by melting (~1290 C) leading to a plasma disruption, ITER FW failure, and TBM FW beryllium- steam reaction producing only 2.2 kg of H 2

17 Summary Idaho National Engineering of significant and Environmental Laboratory results for loss of flow or ex-vessel coolant leak Japanese water-cooled ceramic breeder TBM TBM test cell (a docking cask on VV port ~ 48 m 3 ) would be pressurized to 1.6 MPa, causing cask failure and steam venting into large gallery area (~60,000 m 3 ) If active shutdown does not occur then steam contact with beryllium pebbles could result in TBM FW failure by melting A conservative estimate of H 2 produced needs further analysis

18 Summary Idaho National Engineering of significant and Environmental Laboratory results for loss of flow or ex-vessel coolant leak European water-cooled lithium lead TBM TBM Vault pressurized to 130 kpa FW Beryllium and LiPb reaction with steam produces less than 10 kg of H 2, but further analyses are needed to ensure 2.5 kg limit Russian lithium self-cooled TBM Since all Li-containing TBM structures are placed within the VV port extension, ex-vessel leak means a loss of organic coolant from an intermediate heat exchanger The organic coolant is below auto-ignition temperatures and leaks into a sealed area within the VV (the TBM port transporter) This accident is bounded by in-vessel leak results

19 What Idaho National do Engineering These and Environmental Safety Laboratory Results Mean to US Proposal Team TBMs? Most of the breeding and cooling materials being considered by US PT (excluding molten salts) have already been analyzed for existing ITER TBMs and the results are: Use of liquid lithium requires not only a limited inventory of 35 liters, but all lithium must reside inside of the primary confinement barrier (ITER VV) resulting in the need for an intermediate heat transport loop Use of LiPB requires a limit of 0.28 m 3 inside of the ITER VV at any one time, and fast acting valves that isolate the LiPb inventory outside of the VV during in-vessel leaks must be included in the design The use of helium and water as coolant requires isolation valves to protect the tritium separation systems and pressure relief systems (passive or active) to minimize TBM, TBM vault, and ITER VV pressurization Not all safety issues for already proposed TBMs have been resolved for ITER

20 What Idaho National do Engineering These and Safety Environmental Results Laboratory Mean to US Proposal Team TBMs? (cont.) Insight into self-cooled molten salt TBMs response can be gained from these results, which are: Because Flinabe or Flibe is a low vapor pressure breeder there should not be pressurization concerns during in-vessel or ex-vessel leaks, and thus any mobilized radioactivity salts and ferritic steel material would be confined by confinement barriers Unlike liquid lithium or LiPb which have chemical reaction concerns, the quantity of Flinabe/Flibe will not be limited or need to reside entirely within the ITER VV In-module leaks have no meaning for self-cooled TBMs and there will be no need for isolation systems for a tritium extraction system Safety approach for a self-cooled molten salt TBM should be simpler than for other TBMs

21 What Idaho National do Engineering APEX and Environmental Safety Laboratory Results Mean to US Proposal Team TBMs? Flinabe/Flibe breeders and ferritic steel are not high decay heat materials, implying that decay heat will not be any more of a safety issue than in existing TBMs Flinabe/Flibe ferritic steel TBMs should meet low level waste burial disposal limits Even though tritium permeation and inventory were concerns for an APEX blanket (requiring either a helium cover gas system, or an intermediate heat transfer loop), it may not be a problem for a TBM because of the size of the TBM (i.e. low production and inventories) If Pb is used as a multiplier, then removal of Po-210 and Hg-203 will be an issue that must be addressed Future safety studies of US PT TBMs should prove these conclusions accurate

22 Idaho National Engineering and Environmental Laboratory Conclusions Strong incentive to keep safety approach as simple as practical There are still unresolved safety issues with already proposed TBM material combinations, such as TBMs that have LiPb and beryllium pebble steam reactions Completing a safety dossier for a given TBM is only the first hurdle to clear, there is still the possibility that regulatory approval will be needed for proposed TBMs