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.