TRITIUM ISSUES AT JET

A C Bell
EURATOM/UKAEA Fusion Association
Culham Science Centre, Abingdon
Oxon, UK
+44 (0)1235 464746

INTRODUCTION
The JET machine was originally planned to be capableof operation with DT plasmas with neutron production up to
10 x1023 per year and the torus was therfore constructed withina massive concrete biological shield wall. As the experimental
programme developed, numerous enhancements were madeto the machine and its ancillary systems. Many of these wereassociated with tritium operation such as the construction ofa plant to supply tritium and recycle it for re-use (The Active
Gas Handling System, AGHS) and facilities to enable
maintenance and modification to be carried out remotely.
Tritium experiments took place in 1991 and 1997 andrequired additional measures to be put in place to control
radiological hazards and to handle tritiated waste.
Following these experiments, tritium retention in JET
machine and other components and materials has been an
important consideration and detritiation has been a key issue
for study.

Figure 1: Initial Vessel Configuration

The following sections examine these issues.
TECHNICAL ISSUES RELATED TO TRITIUMOPERATION OF JET
The JET machine has been continually developed andenhanced since its commissioning in 1984 to meet therequirements of the European programme of fusion energy
research. In particular the interior of the vessel has evolved
from an inconel first wall to the current arrangements with anumber of additional heating systems, diagnostics, andpumped divertor with first wall cladding of mainly carbon/
CFC tiles onto which beryllium is evaporated.

Figure 2: Current Vessel Configuration

These changes to the machine have been carried out either
manually by personnel in suitable protective equipment,
which since 1989 has been airline fed plastic suits asprotection against tritium and beryllium, or by remotehandling means involving remote master-slave manipulatorsand robotic arms. Because tritiated dust is produced in theform of co-deposited carbon during plasma operation,
facilities have been built which dock onto the vessel toprovide contamination control and enable waste to be safely
consigned out.

EnclosureContainer withflat bed trolleys

Torus entry port

JG98.368/1c

Short boom, end
effector and tine

Posting platform
Stillage store

Figure 5: Tile Carrier Transfer Facility

Figure 3: Airline Suit Work

The first tritium experiment in 1991, known as the PTE used
0.1 grams of tritium in a once-through mode. Tritium was
supplied to the neutral beam injection system directly from auranium bed and hydrogen isotopes were stripped fromexhaust gases by another uranium bed after initial separation
on a liquid helium cooled finger. The systems used for thisexperiment have been decommissioned.
Because of the significantly large amount involved
recycling of tritium was required for the series of experiments
carried out in 1997 (DTE1) and the more extensive
programme planned at the end of the project (DTE2). The
Active Gas Handling System was designed to cryogenicallyseparate the torus exhaust gas to permit mixed hydrogenisotopes to be separated by one of two isotope separation
systems and impurities by an oxidation/reduction process.
The displacement gas chromatography system was used forenrichment of tritium to 99.5% purity from mixtures withtritium content greater than 1.5%. Although the cryogenic
distillation system is capable of enriching to high purity, thiswould have required a significant inventory in the last columnof > 35 grams of tritium so it has only been used for strippingof tritium from deuterium and hydrogen streams with a DFof >10000.
ML2
ML1

MechanicalForevacuum(MF)

Monitoring

Torus Hall

ExhaustDetritiation
(ED)

CR

NIB 4

TORUS

NIB 8
T2
GIM 15

D2
TDGIS
T2

D2

TDGIS
T2

D2

D2

Gas Distribution
Box (GD)

Intermediate Storage(IS)

Isotope
Separation
Gas chromatography
(GC)
D2

Gas
Introduction
box (GI)

D2
T2

Isotope
Separation
Cryodistillation
(CD)
D
T2 2

T2

Product
Storage(PS)

T2
Torus Basement

Figure 4: Remote Handling Manipulator

H2 (D2)

Active Gas Handling System (AGHS)

Figure 6: Active Gas Handling System (AGHS)

JG00.34/1c

O
RY

Analytical
Laboratory
(AN)

T2

Amersham U–bed

YOPUMP
CR

YOPUM
CR
P

C

NIB Auxiliary ventilation

Impurity
Processing(IP)

Cryogenic
Forevacuum
(CF)

YO

NIB Crown
Torus Crown

Diagnostics

Cooling gas(N2) stream

CF 100Valve(Gas outlet)

JG95.210/6c

Thermocouples

10

JET(1997)

5

Thermocoaxial
heaterValve(Gas outlet)

Chambers
Tubular filter
Uranium powder
Inner cylinder

TFTR(1994)

JET(1991)

00

JG97.509/7c

Process containment
CF 100
Second containment
Ni Fins

JET(1997)

15

Fusion Power (MW)The AGHS uses uranium beds for transfer and storage ofhydrogen isotopes, this choice having been determinedamongst other things by the need to avoid disproportionation
effects. These beds have successfully handled several tensof m3 of hydrogen isotopes and continue to exhibit the original
pumping/storage characteristics.

1.0

2.0

Figure 7: JET Uranium Bed

Drain

5.0

6.0

Figure 9: Fusion Power during JET tritium experiments

was precluded. The in-vessel work in the shutdown following
DTE1 was therefore planned to be carried out completely
using remote handling techniques. The main task of changing
the divertor tiles required an additional boom to be used toenable tiles to be detached and transferred out of the vessel.

Recombiner
500°C

HeaterHeater on

VA-8301A

Molecular
sieve

Drain

VA-8213A Blower

Condenser

Drain

VA-8313A Blower

Heater off

VA-8101B

VA-8201B

VA-8301B

Drier 1(adsorption)

Drier 2(regeneration)

Drier 2(stand by)

JG98.465/2c

Heater off

VA-8201A

Drain

VA-8113A Blower

Condenser
Molecularsieve

VA-8101A

4.0

Recombiner

Condenser
Molecular
sieve1) Secondary containments AGHS
2) Elephant trunk in AGHS
3) Emergency room detritiation
4) Emergency/planned ventillationand detritiation of Torus/MBS
5) AGHS detritiated impurities

Drain
Condenser

Heat
exchanger

Heat
exchanger

Filter Condenser

Recombiner
100°C

A key system was the Exhaust Detritiation System (EDS),
which, as well as providing depression for the containment
boxes of the AGHS, provided detritiation of air streams during
maintenance (and also emergency detritiation in the event ofan up-to air accident in the torus).

3.0Time (s)Figure 8: Exhaust Detritiation System

The AGHS supplied about 100g during DTE1 split about
35g to the vessel by direct gas injection and 65g to the Neutral
Beam injection system, of which 0.6g were injected.
During DTE1, fusion power of 16 MW was produced forabout 1 second. However longer pulses were also obtained
which required loading of the wall with tritium. These andother transport experiments in tritium plasmas meant that at
one point up 11.5g of the site tritium inventory of 20g was
retained in the machine.
A significant consequence of DT operation was increased
activation of the JET inconel vacuum vessel. Prior to this,the maximum radiation level in the JET vessel was a few
hundred microSv/hr, decaying with a 70 day half life frommainly Co58. This permitted man-access to the vessel, inpressurised suits to protect against the Be hazard. Following
DTE1, the dose rate was initially 6 mSv/hr and man-accessFigure 10: Double boom arrangement for tile exchange

TRITIUM RETENTION
Predictions based on the PTE would have suggested much
lower levels of retention. However the configuration of themachine had been altered significantly since the PTE by theaddition of the pumped divertor in the lower part of the vessel.
Co-deposition of carbon deuterium material in the divertor had
been observed in earlier DD operation with the new configuration
and a similar effect was expected with DT operation.

12
CycloneTritium Inventory (g)108Cleaning
head

6

4

0
41400

JG02.30-1c

2
JET Inventory
PTE - based prediction
42400
43400
44400
Pulse Number
Figure 11: Tritium Retention

Ball valves

Cyclone pot

The material was in the form of extremely friable flakes
which readily disintegrate into dust in the 10 micron range.
A characteristic of this dust which became apparent during
its sampling and the handling of divertor components
contaminated by it was the ease with which tritium transferredfrom it to other materials such as PVC leading to bulk aswell as surface contamination. This is now a significant waste
issue in the JET facilities.

JG02.30-2c

CFC tiles

Outerlouvres

JG98.387/4c

Figure 13: Cyclone Pot Cleaner

Cryopump

Innerlouvres

Vacuum cleaner
body

Figure 12: Cross-section of Divertor

Preparations were therefore made to vacuum clean thedivertor in the shutdown following DTE1 to enable thecodeposited material to be collected for analysis. This wasdone using remote handling techniques. The collectedmaterial was transferred in cyclone pots to a glove box in the
AGHS to enable the inventory to be measured and samples
taken. From the earlier analysis of DD material, the tritium/
carbon ratio was expected to be high and stringent precautions
to prevent contamination spread and to handle the high tritium
offgassing rates were employed. Calorimetry was employed
to determine the inventory of 1.47g of material. The tritium
concentration was ~1.5TBq/g and offgassing rate of ~3GBq/hper gram of material.

Figure 14: AGHS Sampling Glovebox

At the start of the shutdown, the in-vessel inventory had reduced
to 6g. With the removal of the dust samples and divertor
components, together with the effect of the air purge, it had
reduced to 4g by the end of the shutdown. Figure 16 shows thecontinued reduction of the succeeding period of plasma
operation. It can be seen that a plateau of around 1.7g is reachedTable I: Tritium Recovered from Pumped Divertor Regenerations
Year
Days
Total Gas Total Tritium Average Tritium
No ofTritium
Regenerations Operations Pumped Pumped Concentration Recovered
bar litrebar litrepercent
mg/day
1998

40

91

1566

0.31

0.0198

0.843

1999

64

159

1936

0.19

0.009

0.296

2000

53

150

2709

0.12

0.004

0.198

2001

42

61

2045

0.04

0.002

0.162
JG01.495-3c

Table II: In vessel tritium concentration compared to vessel
tritium inventory and water collected in EDS during shutdown.
Shutdown

Average Tritium
Concentrationin Vessel
MBq/m3

Tritium
in
Vessel(MIP) g

Tritium
collected attritiatedwater gRTE
(1998)

117

6.2

2

Figure 15: SEM of tritiated dust

PIT
(1999)

55

2.7

0.27

by mid 2000. This was postulated to be associated with dust andflakes which had been deposited on divertor components andthen spalled off into the void below the divertor. An exercise
was undertaken during the 2001 shutdown to gain remote accessto the area and recover a proportion of the material. A significant
quantity of dust was observed and samples are being analysed.

SRP(2001)

23

1.6

Ongoing

Tritium external to AGHS
4

JG01.495-4c

TRITIUM CONTENT OF OTHER MATERIALSAs well as components within the vacuum vessel and
AGHS, other machine components such as the neutral beam
central column were exposed to tritium during DTE1.
Handling of these large components has required special
arrangements for control of contamination to enable transferfrom the machine.

Tritium g

3

2

0
01/01/1998

JG01.495-5c

1

27/09/2000

16/05/1999

09/02/2002

Date
Figure 16: Tritium inventory during plasma operation

Recovery of tritium from the machine took two forms.
During plasma operations, exhaust gas was processed andtritium was separated for reuse. Table 1 shows the gas
separated from the regeneration of the cryogenic divertor
pump, which traps most of the hydrogen isotopes during
plasma operation. During shutdowns, tritium from ventilationof the vessel (and other areas) is collected by the EDS in theform of tritiated water. Table 2 shows the amounts collected.

Figure 17: Neutral Beam central column

In addition, with the vacuum vessel operating at around 300oC
during DTE 1, permeation of tritium through thin the bellows
section of the vacuum vessel into the helium filled interspace
and thence into the torus hall atmosphere lead to measurable
tritium in the torus hall and tritium uptake by a range ofmaterials.
160

HTO
HT

80

JG01.485-2c

40

02/02/98

05/01/98

08/12/97

10/11/97

13/10/97

15/09/97

18/08/97

21/07/97

23/06/97

26/05/97

0
28/04/97

kBq/m3120

As discussed earlier, material which is in contact with tritiateddust also becomes bulk contaminated. This affects the rangeof tenting materials and personal protective equipment which
when disposed of may be described generically ashousekeeping waste .
Analysis of the bulk activity of materials below the limit
of detection of calorimetry generally has to be done bydestructive methods which are time consuming andexpensive. It has therefore not yet been possible to analyse
the tritium content of all materials but the values given below
give an indication of the range in the torus, torus hall and
AGHS.

Figure 18: Torus Hall tritium concentration

Concrete
Paint
RubberPlastics
Housekeeping Wastes
MetalsGraphite/CFC tiles
Co-deposited flakes

30 Bq/gup to 300 Bq/g
100 — 2000 Bq/g
10 Bq/g — 20kBq/g
up to 60 kBq/g
0.1 Bq/g — 1MBq/g
up to 140 MBq/g
up to 1.5 TBq/g

Table III
Class

Form

Current Disposal Status

Detritiation and
Related R&D
necessary

Techniques available or underconsideration(Measurement techniques in italics)

Exempt waste

Solid

Free release

Techniques and protocols to demonstrate
clearance to the satisfaction of
the regulators

Improved QA and sampling

Housekeeping
LLW

Plastic sheeting and protective
clothing, paper, filters.

BNFL Drigg

Volume reduction and control of outgassing.

Surface cleaning
Compaction(Outgassing)Housekeeping
ILW

Plastic sheeting and protective
clothing, paper, filters

Package immobilise then long term
storage at UKAEA Harwell until
UK national ILW site available

Reduction to LLW where practicable andmethods of demonstrating this.
Control of outgassing Understanding
mechanism of contamination

Surface cleaning
Moist gas heat treatment
Incineration
14
Ozone treatment(Calorimeter, Outgassing)

Plant materials

Mainly metals exposed at low
temperatures but may include
others such as ceramics

LLW to BNFL Drigg. Storage of
ILW to be held on site pending
resolution of treatment options

Reduction to LLW where practicable andmethods of demonstrating this.

High temperature air purge
Surface reactions to promoterelease of molecular hydrogen
13,16
Flame heating(BIXS)
17(PIN diode)

Metals exposed at hig
temperatures and in vessel
components

Storage on site pending resolutionof treatment options

Techniques for measurement including depth
profile. Bulk detritiation

Vacuum Melting/Sparging
13,15
Purge gas heat treatment
15
Flame heatingHigh specific activity
non-metals

Tiles and associated material
from in-vessel. Activatedcharcoal

Storage on site pending construction
ofrecovery plant

Recovery of tritium and control of outgassing.
Preferable reduction to LLWFlame heating, high temperature
19oxidation and recovery, UV laser
20
YAG laser

Getter Beds

Uranium, ST707, molecular sieve

In useRecovery of tritium and control of out gassing
and chemical reactivity to permit dismantlingand disposal.

Isotopic swamping and controlled
13oxidation

Organic Liquids

Vacuum oils, Liquid Scint. cocktails

Incineration at off-site facility

Methods of dealing with higher activity

On-site oxidation and recovery
Isotopic swamping

High level TritiatedWater

Low volume high chemical purity
from process operations and EDS

Recovery at off-site facility

Water detritiationplant with recovery

Catalytic exchange/Electrolysis

22,23

Low level Tritiated
Water

High volume including otherspecies such as Be

Discharge to River Thames after
filtration

Minimise dilution. Increase atmospheric emissions. Catalytic exchange/Electrolysis
Water detritiation

22,23

Process gas

AGHS

Discharge or treat with EDS

Monitor environmental impact and effectsof HTO, HT and CH3T. Review dilution
mechanisms and ability to recover(Discriminating samplers)

Ventilation air

From contamination controlled
areas

Discharge or treat with EDS in
accordance with BPEO

Monitor environmental impact and effects of
HTO, HT and CH3T. Review flowrates andability to discharge directly

Local air detritiation systems

Building materials

Structual steels and cladding.
Painted materials. ConcreteStore pending waste
minimisation

Detritiate to exempt level where practicable

Surface removal techniques.
Controlled Weathering–
Surface ContaminatedPlant materials
–

13

13,15

18

Bulk Contaminated

21

JG01.485-4c

Although these were sampled at different times and are
therefore not self-consistent, it can be seen that there is awide variation of materials with contamination levels which
require them to be dealt with as radioactive waste of theappropriate category. This is an issue for current JET
operations and enhancements as well as for futuredecommissioning.
DETRITIATION ISSUES
There is considerable incentive both from an operationaland decommissioning cost point of view to detritiate thesematerials. As well as reducing the ultimate disposal cost,
detritiation reduces off-gassing and cross-contamination, thus
reducing the potential personnel exposure and the cost andtime required to provide protective measures.
It is also consistent with ALARA to recycle tritium wherepracticable. This is particularly the case with the high activity
materials and tritiated water in which gram quantities oftritium were retained during DTE1.
A coordinated programme for studies on the detritiation ofdifferent materials, focussed on the use of JET and its
materials, is being established under EFDA. The key elementsof this are shown in Table III.
CONCLUSIONS
1. Tritium operation of the JET machine required additional
facilities to be constructed which have operated
successfully since the DTE1 experiments.
2. A wide range of tritiated materials is produced in the JET
facilities and improved analytical techniques, particularly
for metals, would assist in assessing their tritium content.
3. A range of detritiation processes has been identified asnecessary for continued exploitation of the JET machine.
4. These requirements will also be relevant for futuremachines
ACKNOWLEDGEMENTS
Since the beginning of 2000, JET has been operated by
UKAEA under the European Fusion Development Agreement(EFDA)
This paper draws on work by many other contributors to
the success of the JET programme. As well as current EFDAJET staff, the author wishes to acknowledge the work ofmembers of Fusion Technology and First Wall Divisions
under the former JET Joint Undertaking.

REFERENCES
The main references for this work are to be found in:
1. Various authors, T echnical Aspects of DeuteriumTritium Experiments at JET , Fusion Engineering and
Design, 47, Special Issue, December 1999.
2. P. D. Brennan et al., Operation of the JET Active
Gas Handling System in Support of Torus Operations
1998-1999 , Proc. of SOFE, 1999.
3. P. Edwards et al., Operating Experience of Tile
Carrier Transfer Facility during the JET Remote Tile
Exchange , Proc. SOFT, 1998.
4. B. Patel et al., Radiological Properties of TritiatedDuct and Flakes from the JET Tokamak , Proc. SOFE,
1999.