1. Vessel Vacuum Tritium Permeation & PbLi/Water Safety Issues Paul Humrickhouse INL Fusion Safety Program October 13 th -14 th, 2011 ARIES Meeting Gaithersburg, MD

2. Fusion Safety Program Discuss tritium inventory and permeation safety limits for vacuum vessel cooling water Describe issues with using water and PbLi in the same reactor ARIES-ACT Conclude with a summary and possible tasks for future research Presentation Outline

3. Fusion Safety Program Parameters used for this vacuum vessel (VV) permeation analysis Upstream T 2 pressure 5 Pa and downstream pressure of zero Metal face plate (wall) made of ferritic-martensitic steel (F82H) Surface area equals 10 3 m 2 and thickness of 0.02 m Wall Temperature 200°C Result for F82H is a flux of 7.0x10 -3 Ci/s. If an austenitic steel is used flux is 8.6x10 -5 Ci/s Coolant Simple equilibrium permeation model used for scoping Surface concentrations u – upstream d - downstream Fick’s law of diffusion Vacuum Vessel Permeation

4. Fusion Safety Program 10 -11 10 -10 10 -9 10 -8 10 -7 10 -6 10 -5 10 -4 10 -3 10 -2 10 -1 10 0 0.911.11.21.31.41.51.6 1000 K/T Hydrogen Permeability, cm 3 (STP)/m-s-kPa 1/2 Ta(s) Ti(s) Pd(s) FS(s) High Nickel Alloys Mo(s) Al(s) PbLi( l ) W(s) Permeability 300 Series SS T of 200°C is 2.1 => Hydrogen Behavior in Metals

5. Fusion Safety Program The allowed public dose from routine release of radionuclides into community drinking water is 0.04 mSv/year (40 CFR 141.66), which translates into a tritium concentration of 20,000 pCi/l (2x10 -5 Ci/m 3 ) for drinking water At the estimated permeation rate for F82H, and an assumed total water volume of 500 m 3, the allowed community drinking water limit is exceeded in minutes after T 2 break through 10CFR61.55 states that low level waste (LLW) shall be ≤ 40 Ci/m 3 in an immobilized form, usually concrete. For a concrete water content of 15%, the time to 265 Ci/m 3 for F82H is ~0.6 years (SS316 is ~50 years). Waste volume over 30 years is 1.7x10 5 m 3 for F82H and 3.3x10 3 m 3 for SS316 As an approximate limit for what can be handled safely (operation releases), the Pickering CANDU reactor water was processed once the tritium concentration reached 500 mCi/l (same as Ci/m 3 ) The time to reach this concentration for F82H is ~1.1 years; for SS 316 the time would be ~90 years Probably the best solution is an online detritiation system if the permeation rate is really this high. ITER will have a water detritiation system that processes 60 kg-H2O/hr. For 500 m 3 of water this system will take about ~1 year to process the entire inventory Tritium Limits

6. Fusion Safety Program If water is used as the VV coolant, then design options should be included that prevent the possibility of common mode failures between the VV and PbLi heat transport systems (HTSs) during design basis accidents (DBA), for example pipe-whips produced by a loss-of-coolant-accident in one system failing the second system One suggestion would be enclosures (a strong box) for the PbLi HTS loops. This would have the advantage of preventing during DBA or operating conditions: – Common mode failures – Water and PbLi mixing to produce hydrogen – Release of Po-210 and Hg-203 – Release of T 2 permeating from the PbLi HTS if the enclosure atmosphere is detritiated – Minimizing occupational radiation exposure during maintenance activities on other systems if the enclosure is shield The enclosure should be compact and as close as possible to the reactor to avoid worker exposure from long PbLi HTS piping Of course, this will not prevent PbLi/water interactions during beyond DBA events like the Fukushima earthquake Safety Related to PbLi/Water Issues

7. Fusion Safety Program Safety issues are the hydrogen, heat, pressurization and hazardous chemicals (LiOH) produced by the chemical reaction between the Li in the PbLi and any water that comes into contact with the PbLi Experiments have shown that the amount of hydrogen generated is strongly dependent on the contact mode Injection – pressurized injection of water into liquid metal (LM) Pouring – pouring of LM into water Layered – pouring of water onto LM Pool – steam environment over LM pool Spray – steam environment present during LM spray Safety Issues Related to PbLi/Water Reactions

8. Fusion Safety Program A large body of work has already been accomplished at HEDL (Jeppson), ISPRA (Savatteri), and University of Wisconsin (Corradini), just to name a few in the area of PbLi/Water reaction. An excellent review paper is: M. Corradini, and D. W. Jeppson, “Lithium alloy chemical reactivity with reactor materials: current state of knowledge,” Fusion Engineering and Design, 14 (1991), p. 273-288. Corradini’s summary is that for the layered, pool, and spray contact modes the reaction is self-limiting by the formation of solid LiOH and Li 2 O crusts that shield the LM from the water/steam interface, but other contact modes may be a problem, especially the injection contact mode However even within the various modes of contact much uncertainty exists due to pool geometries, injection velocity, injection direction, PbLi and water temperature, PbLi to water mass ratio, etc. Safety Issues Related to PbLi/Water Reactions (cont.)

9. Fusion Safety Program D. W. Jeppson and C. Savatteri BLAST Experiment (1991) Based on mass spec analysis, ~8% of the lithium in the LM reacted to form H 2 LM temperature Test pressure Injection Mode Geometry (applicable example)

10. Fusion Safety Program In our ARIES-CS safety paper, we examined a LOCA in a PbLi HTS inside a pipe chase. The result was a PbLi pool of ~100 m 3. The volume of Li in this pool would be 17 m 3 If a water pipe at the bottom of this pool were to break due to thermal stresses, according to the data from the BLAST experiment the amount of Li reacted in the pool would be 17 x 0.08 or ~1.4 m 3 Reacting this volume of Li (~750 kg) would produce ~110 kg-H 2. Based on MELCOR calculations preformed by B. Merrill for Unit 1 of Fukushima, about 1000 kg of H2 was produced by Zirc oxidation at the time of confinement building failure Possible Hydrogen Generation Scenario

11. Fusion Safety Program Permeation into the VV cooling water may create a LLW volume concern. An online water detritiation system should be used to minimize water tritium concentration. There is a large uncertainty in the presented numbers due to an assumed upstream tritium pressure. Will investigate methods to more accurately estimate this pressure If Ta is being considered as a divertor material, a simple TMAP model should be used to predict the permeation rate for this divertor The use of water and PbLi in the same reactor always produces safety concerns. It is proposed that the PbLi system be placed in a separate enclosure to eliminate common mode failure accidents. Perhaps a design option is available to minimize PbLi/reactions even under BDBA conditions The INL needs to start developing MELCOR and TMAP models for ARIES ACT. Perhaps code development will progress to the point where these codes are merged, allowing more accurate tritium permeation calculations Summary