TBM WG Meeting UCLA, March 2-4, 2005 Brad Merrill/Fusion Safety Program DCLL TBM Safety Update

Presentation Outline Discuss General Safety Requirements for test blanket modules (TBMs) Present MELCOR model developed to analyze accidents for the dual cooled (DC) TBM Discuss Reference Accidents that must be addressed by the DC TBM Design Description Document (DDD) Describe ITER IT analysis specifications and present MELCOR results for each accident case Present TMAP permeation model developed for the DC TBM and the predicted operational permeation results Summary

General Safety Requirements for TBMs Demonstrate that a pressure transient inside primary and secondary confinement barriers does not exceed ITER design limit of 2 atm –ITER vacuum vessel (VV) and pressure suppression system, and all ex-vessel parts of cooling and other TBM auxiliary systems represent the primary confinement barrier –Test blanket cells and TWCS vault represent 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 –PbLi 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/PbLi interaction should be performed. –A detailed analysis requires that a PbLi spray into water in the VV be considered to be atomized by the helium coolant pressure (resulting in droplets that are ~ 2 mm in radius), making an analysis problematic since reaction rate data does not exist such droplets

General Safety Requirements TBMs (cont.) Chemical reactions (cont.) –Data similar to that needed for this PbLi/H 2 O contact mode is Jeppson’s pouring data (2 g of 600 °C PbLi into 4000 g of 95 °C H 2 O), which gives an initial drop radius of ~8 mm. Jeppson found that only ~50% of the Li reacted (at 50%, the hydrogen produced would be 1.9 kg for the DC TBM total inventory of 0.42 m 3 ). Even for the most severe contact mode (high pressure H 2 O injection into PbLi), the maximum reported reacted Li is not greater than 70 % (at 70%, the hydrogen produced would be 2.5 kg for the DC TBM total inventory). –If not an acceptable solution to the ITER IT, then additional tests may be needed –Beryllium on the first wall (FW) of a TBM 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 –Maximum TBM FW temperature < 350 o C in the post-shutdown period which begins a few tens of minutes after shutdown Helium should be limited to 40 kg to avoid fouling of VV pressure suppression system by the helium

Schematic of MELCOR TBM Model First wall Concentric pipe Permeator PbLi/He HX Back plate He pipes He/H2O HXs Vacuum vessel Be/FS/HE/FS/SiC Drain tank Port cell TWCS vault 30 control volumes 37 flow paths 72 heat structures (psuedo 3D TBM conduction) 6 valves 1 rupture disk 1 pump and 2 circulators

Loop Coolant Temperatures During Pulses Peak temperature of 650 °C requires flow at 6 kg/s, at which temperature 70% of the PbLi heating is lost to the helium by convection and conduction (split is 80% helium vs. 20 % PbLi, which is the reverse of Sergey’s DEMO results). Thermal equilibrium not achieved during a pulse, but temperatures do not ratchet up for repeated pulses Time (s) Temperature (C) PbLi HX PbLi Zone 1 FW He inlet FW He outlet PbLi Zone Time (s) Temperature (C) PbLi Zone 1 (6 kg/s) PbLi Zone 1 (0 kg/s)

Reference Accidents That Must be Analyzed Ex-vessel LOCA analysis to determine: –Pressurization of TBM vault –Behavior of TBM without active plasma shutdown 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 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

TBM Ex-vessel LOCA Specifications Event type Ultimate safety margin Objectives Show that in-vessel hydrogen generation is limited Show that pressure transient inside coolant vault stays within design limits Show how fusion power shutdown affects transient Show that post accident cooling is established to a safe shutdown state Scope of analysis Focus on correct/conservative description of possible chemical reactions of the PbLi with steam (use chemical reaction rates and safety factors called out in SADL) Predict confinement barrier overpressure Initiating event A double ended pipe break in a TBM FW cooling loop is postulated to occur in the largest diameter pipe of the HTS, discharging coolant into the test cell during a plasma burn

TBM Ex-vessel TBM LOCA Pressure Results LOCA assumed to start at the end of a reactor pulse flat top (300 s) Port cell relief valve (set to open at 0.4 atm pressure differential with TWCS vault and to re-seat at 0.01 atm pressure differential) limits test cell pressure to 1.5 atm, not exceeding confinement barrier design limits of 2 atm Time (hr) Pressure (atm) Test cell TWCS vault Vacuum vessel Time (s) Pressure (atm) 2.0 Test cell TWCS vault Vacuum vessel Expanded view

TBM Ex-vessel LOCA FW Temperature TBM FW beryllium evaporates and disrupts plasma ~ 90 seconds after LOCA starts Beryllium ignites on “hot strip” (region with a surface heating of 0.5 MW/m 2 ) after steam enters VV TBM FW steel melting occurs Expanded view Time (hr) Temperature (C) First wall Second wall Time (s) Temperature (C) First wall Second wall

TBM Ex-vessel LOCA Hydrogen Production Complete oxidation of FW beryllium on FW “hot strip” results in 0.12 kg of hydrogen created Other TBM FW regions do not reach temperatures that will ignite the Be prior to reactor shutdown H 2 is below limit of 2.5 kg Time (s) FW beryllium thickness (mm) Be thickness H2 production FW Hydrogen Production (kg)

TBM in Breeder Box LOCA Objectives Assess TBM box pressurization caused by release of TBM coolant Demonstrate that the tritium processing gas system and VV are protected from pressurization Show that decay heat is removed passively Show that no excessive chemical (Li-steam) reactions occur Initiating event A break of the largest pipe inside the TBM is postulated (one FW channel assumed) Transient sequences A one hour loss of offsite site power is assumed to coincide with the initiating event Credit for any active (safety) system can not be taken unless the system is designed to one specific function; for example, emergency pressure relief Fusion power is terminated by melting of FW, inducing a plasma disruption

Time (s) Pressure (atm) TBM in Breeder Box LOCA Pressure Results Pressure Results TBM pressure reaches ~85 atm after FW channel break, while FW failure occurs at ~ 400 s (end of pulse flat top) Test cell and TWCS vault pressure remain below design limit Pressurization of tritium extraction system will not be an issue if the permeator is designed to withstand the FW helium pressure Time (s) Pressure (atm) Blanket FW helium Time (s) Pressure (atm) Blanket Vault Test cell V

TBM in Breeder Box LOCA Oxidation Temperature/Oxidation Results Beryllium oxidation is not self sustaining after FW failure due to PbLi entering FW channels, adding cooling, additional heat capacity and enhanced conduction PbLi hydrogen generation is less of an issue since most of the PbLi moves into the drain tank Time (s) Temperature (C) FW Time (s) Beryllium thickness (mm) Hydrogen generation (kg) Time (s) PbLi volume (m 3 ) Drain tank Loop Hydrogen Beryllium

In-vessel TBM LOCA Objectives Assess VV pressurization caused by release of TBM coolant Show that decay heat is removed passively Show that no excessive chemical (Be-steam) reactions occur Initiating event Multiple breaks of the TBM FW cooling pipes is postulated with a coolant spill into the VV. Assume a double ended rupture of all FW pipes in one 10 cm high toroidal ring. Transient sequences Fusion power is terminated by coolant ingress, producing a plasma disruption and failure of ITER FW The FW of the failed loop is first cooled down with residual cooling from coolant present before complete drainage. After the coolant inventory of the failed loop is lost, the FW of the failed loop will be cooled by steam convection and radiation to the VV. The VV cooling system is assumed to operate in natural circulation mode.

In-vessel TBM LOCA Pressure Results Pressure Results TBM depressurization ~ 2.5 s TBM helium does not foul the VV pressure suppression system, causing only a 5 kPa pressure increase above that of ITER FW steam only (13 kg of the TBM FW loop helium was injected into the VV) Time (s) Pressure (atm) Time (s) Pressure (atm) Time (s) Pressure (atm) VV pressure with TBM helium VV pressure without TBM helium TBM helium VV

In-vessel TBM LOCA Oxidation Temperature/Oxidation Results FW temperatures do not result in a beryllium oxidation thermal runaway; results show a steady decline due to VV steam cooling Hydrogen generation is not an issue FW Time (s) Temperature (C) Time (s) Hydrogen produced (g)

TBM Tritium Permeation Analysis ITER allowed TBM release to the environment is ~ 1 mg-T as HTO/a, assuming 99% efficient cleanup system this translates into an in-building permeation limit of ~ 100 mg-T/a A TMAP model has been developed to examine permeation from the TBM and ancillary system, this model includes: –All of the piping (helium pipes not concentric, 380 °C or 440 °C), heat exchangers, and walls of the TBM (no SiC permeation barrier credit taken in PbLi pipes and permeation in the TBM is only from the gaps) –A vacuum permeator composed of 20 tubes (5 m length, 0.01 m diameter) for low performance operation, and 2 tubes for high performance operation –PbLi mass transport coefficients based on Harriot & Hamilton correlation (K m = 1.2 mm/s in permeator) –Sawan's T 2 production rate of 1.43e-6 g/s applied over 600 s pulses which translates to 2.15 g-T/a for 3000 pulses/a (note that the 100 mg-T/a limit is less than 5% of total)

Concentric pipe (FS walls) Permeator PbLi core PbLi/He HX Non-Hartmann Gaps Hartmann Gaps First wall Second wall Rib walls Back plate First wall He Rib He He pipes (FS walls) He/H2O HXs Schematic of TMAP TBM Model

High Performance TBM Tritium Permeation Results TBM PbLi tritium concentration reaches an oscillatory equilibrium after ~10 pulses, while helium pipe FS walls reach an equilibrium after ~ 70 pulses (2 days) Annual release based on 3000 consecutive pulses is 78 mg-T/a from helium pipes, and 60 mg-T/a from inlet PbLi pipe (total ~140 mg-T/a which is above limit of 100 mg-T/a) Number of pulses T2 pressure above PbLi (Pa) Number of pulses Annual release rate (mg-T/a) He pipe walls PbLi pipe wall

TBM Tritium Permeation Issues With the permeator option we are close to satisfying tritium release limits and could be under the limit if some credit can be taken for SiC insert in PbLi pipe Permeation from the pipes could be reduced if SS316 were used instead of FS because permeation coefficient for SS316 is a factor of 4 lower than FS at 600 °C (works for PbLi pipes but not helium pipes) Baseline TBM conditions give permeation below the limit for Nb permeator tubes (total ~ 45 mg-T/a), but MANET (martensitic steel) will not (over limit by a factor of 3) If a bubble column is used, the permeation rate would scale as the square root of pressure {~ (100/.5) 1/2 or 14 times higher} ITER may require permeation barriers on pipes (alumina coating is a possibility) as a matter of operational safety if releases are near the annual limit and can not be verified during TBM operation

Summary TBM pressurization of the VV, vaults and test cell is within ITER IT acceptance criterion TBM FW beryllium oxidation results in thermal runaway but hydrogen generation does not exceed allowed limit of 2.5 kg PbLi inventory is over the allowed 0.28 m 3, but the argument will be advanced that only 50% of Li can react Helium inventory is estimated to be 17 kg for the FW helium loop, but we will need to know this value accurately because of ITER limit of 40 kg Tritium permeation may require a permeation barrier for piping to meet ITER limits Most of the required analyses for the DDD have been completed, but parameter studies for the in-vessel LOCA accident are still needed