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Safety Analysis of the US Dual Coolant Liquid Lead Lithium ITER Test Blanket Module
Brad Merrill (INL), Susana Reyes (LLNL), Mohamed Sawan (UW-M), and Clement Wong (GA)
8th IAEA Technical Meeting on "Fusion Power Plant Safety", Vienna, 10-13 July, 2006
Presentation Outline
• Technical description of US Dual Coolant Liquid Lead Lithium (DCLL) Test Blanket Module (TBM) concept
• TBM and ancillary system inventories and materials• Safety Analyses
– Normal releases– Accident analyses that appear in the TBM Design Description
Document– Radioactive waste estimates
• Future activities required by ITER– Failure Modes and Effects Analysis (FMEA), additional Accident
Analyses, and Occupational Radiation Exposure (ORE) Analysis
US DCLL TBM is a DEMO Prototype
DEMO DCLL Blanket US ITER DCLL TBM
PbLi Inlet/Outlet Pipe
PbLi Flow Channels
Back Plate
He Inlet/Outlet
Pipe
First Wall
He out
He in
PbLi flow
PbLi outPbLi in
Back
2mm Be front face
FCI is the Thermal and MHD Insulator lining all PbLi channels
He out
He in
All FS structures are He-cooled @ 8 MPaPbLi self-cooled flows in poloidal direction
PbLi inPbLi out
FW He counter flow
PbLi flow
US DCLL TBM Schematic
Mid-plane cross-section
DCLL TBM PbLi Ancillary System in TransporterITER
Transporter
Dual PbLipipe
Drain tank PbLi primary heatexchanger
Pipe chase view
PbLi tritium extractor(vacuum permeator)Surge tank
F, Ci F, Ci
Gtot
z
∆z
G
PbLi, T
T2
PbLi, T
FS membrane
PbLi pump
US DCLL TBM and Ancillary System Layout
TWCSVault
ITERShield
ITERVV Pipe chase
(five or six pipes depending ontype of tritium extraction system)
Five lines into TBM• Concentric PbLi pipe• FW helium inlet/outlet pipes• TBM vent/drain pipes
US DCLL TBM Materials & Thermal ParametersComponent Materials Dimensions
Module
Structural material Ferritic steel, e.g. F82H
Breeding material PbLi V=0.28 m3
FW/structural coolant 8 MPa helium
Flow channel insert SiCf /SiC
FW coating 2 mm Be M=4.6 kg
PbLi Loop
Breeding material PbLi V=0.12 m3
Piping (concentric) Ferritic steel L=8 m, D=90/160 mm
Intermediate heat exchanger Ferritic steel A=4.2 m2
Permeator Ferritic steel A=3.1 m2
FW Helium Loop
Coolant 8 MPa helium 18 m3
Piping (2 pipes) Austenitic steel L=95 m, D=90 mm
Water heat exchanger Aluminum A=1 m2
Intermediate Helium Loop
Coolant 8 MPa helium 17 m3
Piping (2 pipes) Stainless steel L=95 m, D=90 mm
Water heat exchanger Aluminum A=1 m2
Thermal hydraulic parameters
Module thermal power for 600 s, MW
1.375
FW/structure primary He loop
Module power fraction 0.54
He Tin/Tout, C 300/440
He system pressure, MPa 7/8
He mass flow rate, kg/s 0.649
Pb-17Li loop
Module power fraction 0.46
PbLi pressure, MPa 0.2/2
TBM PbLi Tin/Tout, C 300/470-650
PbLi mass flow rate, kg/s 72
Intermediate helium loop
He pressure, MPa 8
Tin/Tout, C 180/300
He mass flow rate, kg/s 0.64
Radioactive Source Inventories
Source Inventory 1Dose (mSv)
Tritium
Structure 235 mg
PbLi 2 mg
FW Be 33 mg
Breeding Material
Po-210 1.8 Ci 0.08
Hg-203 36 Ci 0.002
TBM structure
Total Oxidation in steam
6x10-3/day
Ta-183 69%
W-187 14%
Co-60 7%
Mn-54 3%
1600 times < ITER
1Typical weather conditions – Pasquill-Gifford D, 4 m/s
Nuclear and Chemical Energy SourcesNuclear energy sources (MJ)
1Beryllium clad (MJ)Plasma disruption (1.8 MJ/m2 on a 1.25 m2
surface)2.25
Delayed plasma shutdown (normal: 3 s delay, 1 s ramp-down)
4.8
Decay heat integrated over:1 minute1 hour1 day 1 month
0.2121361640
Chemical reaction energy 330
Hydrogen generation 1.2 kg
1ITER Safety Analysis Data List reaction rates in MELCOR
Breeding material (MJ)
Latent 380
Chemical 115
Hydrogen(Highest reported lithium reaction)
2Contact mode: High pressure water injection = 71.5% 3Pouring mode: 20 g of 600 C PbLi poured into 4000 g of 95 C water = 50%
2.55 kg
2.0 kg
2M. Corradini, and D. W. Jeppson, Fusion Engineering and Design, 14 (1991), p. 273-288, L. S. Nelson, University of Wisconsin (UWFDM-1031, 1996)3D. W. Jeppson, Nuclear Technology/Fusion, 4 (1983), p. 277-287.
DCLL TBM Tritium Operational Release Analysis
• ITER allowable TBM release to the environment is ~ 1 mg-T as HTO per year, assuming a 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 credit taken for SiC as a permeation barrier in PbLi pipes and permeation in the TBM is only from the gaps between SiC insert and FS walls)
– A vacuum permeator composed of 20 tubes (5 m length, 1 cm diameter, 0.5 mm thick)
– Assumed efficiency of the helium coolant tritium cleanup system is 95% for a 1% slipstream
– Tritium production rate of 1.59x10-6 g/s applied over 600 s pulses (400 s flat top) which translates to 2.15 g-T/a for 3000 pulses/a (note that the 100 mg-T/a operational release limit is less than 5% of total produced by TBM)
– Pulse scenario of 1 week continuous pulsed operation (600 s pulse with 1800 s dwell) and 3 weeks down time repeated throughout the year for a total of 3000 pulses
Concentric pipe(FS walls)
PermeatorPbLi core
PbLi/He HX
Non-Hartmann Gaps
Hartmann Gaps
First wall
Second wall
Rib walls
Back plate
First wall He
Rib He
He/H2O HXs
Schematic of TMAP TBM Model
Tritium cleanupsystem
He pipes(FS walls)
DCLL TBM Operational Tritium Release• Helium pipe annual release is 160 mg-T/a; PbLi pipe annual release is 135 mg-T/a; Total
~295 mg-T/a which is above the operational release limit of 100 mg-T/a
• If the length of the permeator is increased to 10 m, the tot drops to 245 mg-T/a, and if credit is taken for the SiC PbLi insert to only 145 mg-T/a
• Experiments are planned to investigate permeator performance but we are considering aluminum shrouds for the transporter and pipes as a fall back option
0 100 200 300 400Time (d)H
eliu
m p
ipe
perm
eatio
n ra
te (m
g-T/
s)
0
1x10-5
2x10-5
3x10-5
PbL
i per
mea
tion
rate
(mg-
T/s)
0 100 200 300 400Time (d)
0
1x10-5
2x10-5
3x10-5
Reference Accidents Analyzed• 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
• Complete loss of TBM active cooling
MELCOR
Code analyzed with
CHEMCON
MELCOR Model for Reference Accident Analyses
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
Tokamak Cooling Water System (TWCS) vault
• 30 control volumes• 37 flow paths• 72 heat structures
(pseudo 3D TBM conduction)
• 6 valves• 1 rupture disk• 1 pump and 2 circulators
PbLi system rupture disk opens at 4.0 atmand drain tank relief at 3.0 atm
Port cell relief vent opens on a 0.4 atmpressure differential with TCWS vault and reseats at 0.015 atm
Operational Response of TBM Coolant Temperatures During Pulses
• Thermal equilibrium not achieved during first pulse from “hot standby” conditions, but a pulsed equilibrium does develop after several repeated pulses
0 2000 4000 6000 8000Time (s)
300
400
500
600
700Te
mpe
ratu
re (C
)
PbLi HX
PbLi Zone 1
FW He inlet FW He outlet
PbLi Zone 2
TBM Ex-vessel Helium 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 helium cooling loop is postulated to occur in the
largest diameter pipe of the HTS, discharging coolant into the test cell during a plasma burn. This event causes the TBM FW Be to melt, inducing a plasma disruption that fails the ITER FW
• Variant case: Induced disruption damages TBM box and spills PbLi into VV (10 cm2 break)
TBM Ex-vessel TBM LOCA Pressure and Mass Results
• LOCA assumed to start at 100 s before the end of a reactor pulse flat top (at 300 s of burn)
• 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
0.0 0.5 1.0 1.5 2.0Time (hr)
0.0
0.5
1.0
1.5
2.0
Pre
ssur
e (a
tm)
Test cellTWCS vaultVV
2600 2800 3000 3200Time (s)
0.8
1.0
1.2
1.4
1.6P
ress
ure
(atm
)
2.0Expanded view
Test cell
Vault
0.0 0.5 1.0 1.5 2.0Time (hr)
0
5
10
15
Hel
ium
mas
s (k
g)
0
10
20
30
40
50
Ste
am m
ass
(kg)
Test cell
Vault
0.0 0.5 1.0 1.5 2.0Time (hr)
TBM Ex-vessel LOCA FW Temperature
• TBM FW beryllium evaporates and disrupts plasma ~ 90 seconds after LOCA starts
• Beryllium on “hot strip” (mid-plane of TBM FW) does not ignite after steam enters VV but does lose half of its beryllium by oxidation; and the total hydrogen generation for the FW is 0.15 kg
Expanded view
0
250
500
750
1000
1250
1500
Tem
pera
ture
(C)
0.0 0.5 1.0 1.5 2.0Time (hr)
FW
SW 2800 3000 3200Time (s)
0
500
1000
1500
FW
SW
0
1
2
3
FW b
eryl
lium
thic
knes
s (m
m)
FW
.00
.05
.10
.15
FW H
ydro
gen
Pro
duct
ion
(kg)
0.0 0.5 1.0 1.5 2.0Time (hr)
FW hot strip
TBM Ex-vessel LOCA with Simultaneous Spill of PbLi into Vacuum Vessel
0.0 0.5 1.0 1.5 2.0Time (hr)
0
500
1000
1500
Tem
pera
ture
(C)
Base case
Variant case0
10
20
30
Pre
ssur
e (a
tm)
0.0 0.5 1.0 1.5 2.0Time (hr)
Base case
Variant case
Break quickly de-pressurizes breeder zoneFW Temperature is lower after plasma disruption
Vol
ume
(m3 )
0.0
0.1
0.2
0.3
0.4
0.5
0.0 0.5 1.0 1.5 2.0Time (hr)
TBM system
VV
2800 2825 2850 2875 2900Time (s)
1.0
1.5
2.0
Pre
ssur
e (a
tm)
Base case
Variant case
VV pressure increase by PbLi interaction is ~10 kPaPbLi volume in VV is 0.3 m3, giving < 2.5 kg H2
DCLL TBM Radioactive Waste Assessment• The radwaste classification was evaluated according to the US Nuclear Regulatory Commission
(NRC) 10FR61 and Steve Fetter’s fusion waste disposal rating (WDR) concentration limits. The limits are based on the assumption that all solid components are crushed before being disposed (no voids)
• Although the Fetter limits are generally more conservative, the WDR are much lower than unity and therefore qualify for shallow or Class C land burial
Structure Primary Contributor
Half life (y) NRC WDR Fetter WDR
F82H Nb-94 2.03×104 6.9×10-3 1.3×10-2
Mn-53 3.70×106
Ni-59 7.50×104
Nb-91 6.80×102
Pb-17Li Pb-205 1.52×107 2.9×10-9 8.7×10-3
SiC C-14 5.73x103 7.3×10-14 2.1×10-4
Be-10 1.51×106
TBM 6.9x10-3 2.2x10-2
Future TBM Activities Required by ITER• The ITER International Team has asked TBM Parties to provide, by January 2007, TBM
safety assessments that will be included in the safety files for ITER’s Report on Preliminary Safety (RPrS) that is required for a License to Construct. The input for the DCLL TBM and ancillary systems must include:
– Technical description,
– Source terms (radioactive, energy, and chemical)
– Operational releases
– Plant worker occupational radiation exposure (ORE) estimates
– Failure modes and effects analysis (FMEA) study
– Consequence analysis of selected design basis and beyond design basis accidents
– Waste disposal analysis
• Most of this input already exists in the TBM safety assessment already contained within the DCLL TBM Design Description Document, but new accident analyses may be identified by the FMEA. An ORE analysis is also needed.
• Beyond the RPrS, the ITER IT is requesting a safety assessment that covers these same safety areas, but in more depth. This assessment, TBM Dossier on Safety (DOS), will be incorporated in ITER’s Final Safety Report (FSR) submittal to obtain an Operating License prior to DT plasma operation.