8.  Tritium System Requirements

**5.8.1 Introduction**

The tritium system is a key system for FIRE operations, as well as
providing development information and operating experience for the
fusion program. The development of a safe, low-inventory tritium system
is an important project goal.

2.  **Tritium Injection**

Requirements

Pellet injection will be the primary plasma system for fueling the core
of the FIRE plasma utilizing both high speed pellets and guided lower
speed pellets. Gas injection systems will also be provided for edge
fueling. The total number of tritons in the nominal FIRE plasma is:

Ne = \<ne\> Vp /2 \~ 2.25 x 1020 m-3 x 18 m3 \~ 5 x 1021 tritons

The particle confinement in tokamaks is described by D \~ or p \~ E
which translates into p \~ 0.5 to 0.8 s in FIRE. A fueling rate of \~
0.5 - 1 x 1022 particles/s would be required to sustain the density with
zero recycling. The standard assumption for FIRE and ITER-RC is that He
\~ 5 E which suggests an 80% recycling of helium. We make the additional
reasonable assumption that the same recycling applies to the hydrogenic
species. Therefore, a net rate \~ 0.1 - 0.2 x 10\^22 particles/sec would
be required to sustain the nominal FIRE plasma.

In present experiments with outer mid-plane pellet fueling, the
efficiency is low \~20%. FIRE will be employing vertical launch of
high-speed pellets aimed inside the magnetic axis or slower pellets
guided by tubes to near the inside mid-plane. The injection geometry
will be updated, as more information becomes available from ongoing
experiments. A pellet fueling efficiency of 50% is assumed for FIRE. The
gross tritium fueling rate for the plasma core is then \~ 0.2 - 0.4 x
1022 particles/s.

3.  **Requirements for Potential**

#  Pulse Sequences

The total number of injected tritons required for various scenarios
involving 10 second long pulses is:

2 - 4 x 1022 T/pulse ; 0.8 - 1.6 kCi (\~0.1g) / pulse; 10 s pulse

2 - 4 x 1023 T/day ; 8 - 16 kCi (\~1g)/day; 10 pulses/day

1 - 2 x 1024 T/week; 40 - 80 kCi(\~5g)/week; 50 pulses/week

(where 2.09 x 1019 T atoms = 1 Ci, 104 Ci = 1 g, pulses are 10 seconds
long)

The total number of DT pulses in FIRE is limited to \< 5TJ of fusion
energy, or 2,500 pulses at 200 MW for 10 s, or a tritium fueling
throughput of 2 - 4 MCi. There will also be many partial power pulses
that will consume tritium while not producing optimal fusion power,
therefore the lifetime throughput of tritium is assumed to be increased
by \~ 5 to a total throughput of 10 - 20 MCi. Assuming that this program
takes place over 5 years would require a tritium throughput capability
of 2 - 4 MCi/year.

The fractional tritium burn-up of \~ 5% does not affect these estimates
significantly.

4.  **Tritium Retention and Inventory**

The provisional limit for the tritium site inventory has been set at ≤30
g (\~0.3 MCi). According to DOE STD 1027, FIRE would be classified as a
Low Hazard Nuclear Facility. Similar to the TFTR tritium management
strategy, the maximum tritium in any on-site tritium system is set at 15
g, 1/2 the site limit. The D-T experiments on TFTR and JET have shown
that the use of carbon plasma facing components produced an effective
tritium retention rate of \~ 40%. Assuming the annual tritium throughput
of 2 - 4 MiCi, this level of retention would cause the FIRE system limit
to be exceeded in 1 to 2 months. Therefore, the use of carbon will not
be allowed in the FIRE vacuum vessel. The initial materials for plasma
facing components and divertor plates will be Be and W.

Cryogenic pumps will be installed in the behind the divertors of FIRE to
provide adequate pumping during the pulse. Essentially all of the
injected tritium will end up on the cryopumps. The schedule for
regenerating the cryopumps will be determined to maintain the tritium
inventory \< 30g. The tritium inventory for several regeneration
schedules is:

1\. weekly regeneration: \< 5g of tritium on the cryopumps would be
transferred to the tritium handling system. The tritium separation
systems described below would be able to easily separate the tritium
from the deuterium and other exhaust gases in \< 2 days, so that tritium
could be ready for experiments the following week. Need to estimate the
number of deuterium only shots in a run sequence to estimate the total
number of torr-liters of gas on the cryopumps to see if this is a
reasonable sequence.

2\. daily regeneration: \< 1g inventory on the pumps, probably not worth
the hassle of the regeneration procedure.

3\. monthly regeneration: \< 20g of tritium on the cryopumps if running
continuously with tritium. This level is also expected to satisfy the
explosive limits.

A monthly regeneration of the divertor cryopumps would fit naturally
with the anticipated experimental schedule.

**5.8.5 Tritium Systems for FIRE**

The tritium systems will be similar to those used successfully at TFTR,
and will include Tritium Storage and Delivery, plasma exhaust cleanup,
tritium purification system (for reprocessing the on-site inventory),
appropriate room air cleanup systems, tritium exhaust gas processing
systems, and tritium monitoring for process control and personnel
protection. The block diagram for the tritium system is shown in Fig
5.8.5-1.

The FIRE tritium delivery system will be capable of supplying tritium
with a purity \> 98%. Tritium will be received from a DOE supplier in
hydride transport vessels (HTVs) in quantities up to 25 grams. Tritium
inventory will be loaded into the tritium storage and delivery system
(TSDS) and will be available upon demand (within 6 hours of when
required). The FIRE TSDS will be capable of supplying quantities of
tritium up to 3 kCi per pulse via direct gas injection. This capability
could be upgraded for the long pulse (\~ 40s) pulses in the advanced
tokamak phase.

FIRE exhaust gas will be collected in a plasma exhaust tank where it
will be stored until processed by the on-site tritium purification
system (cryogenic distillation). On-site tritium processing will
separate non hydrogen isotopes from the plasma exhaust effluent and
cryogenically separate tritium from deuterium and protium, thus
producing tritium with a purity of \> 98 % purity. Plasma exhaust
processing will require \~ 24 hours to be recycled back to the tritium
storage and delivery system.

The on-site tritium purification system for FIRE will have a resident
tritium inventory of \~ 10 grams of tritium with a throughput of (up to)
50 kCi (5 grams) / day. ITER had planned to reprocess 164 g of tritium
during the 40 minute cycle period for an ITER pulse\[1\].

Tritium residual gases (in glove boxes and in other small volumes) will
be processed, oxidized, and deposited on disposal molecular sieve beds
for disposal at an off-site facility.

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height="4.036111111111111in"}

5.8.6 Options to the Tritium Inventory

The tritium inventory has been set at 30g (\~0.3 MCi), to allow
sufficient operational flexibility without introducing additional
restrictions. However, there is the potential for reducing the inventory
to even lower levels. If a tritium reprocessing system is used which is
able to recycle the working tritium on a daily basis, then the daily
working inventory is = 20 kCi (2g).

As noted above, ITER was planning on reprocessing tritium at the rate of
\>4 g/minute. If FIRE had a system capable of processing 1g/120 minutes,
then the working inventory could be reduced by an order of magnitude to
2kCi(0.2g). The main contributions to the inventory would now be in
residual holdup in various systems including the vacuum vessel. There
should be a follow-up study to look at the minimum tritium inventory
case.

\[1\] D. K. Murdoch, "Tritium Inventory Issues for Future Reactors";
Choices, Parameters, Limits. Proc., SOFT 1998