- Email: [email protected]
The tritium cleanup experiment in JET
jnurnalnf nuclear materials
Journal of Nuclear Materials 196-198 (1992) 143-148 North-Holland
The tritium cleanup experiment in JET P. Andrew, C.J. Caldwell-Nichols, J.P. Coad, K.J. Dietz, J. Ehrenberg, D.H.J. Goodall, J. Hemmerich, L. Horton, J. How, R. Lasser, P. Lomas, M. Loughlin, G.M. McCracken, A. Peacock, G. Saibene, R. Sartori, P. Thomas and T. Winkel JET Joint Undertaking, Abingdon, OxJordshire 0)(14 3EA, United Kingdom
The first tritium fuelled discharges in JET were followed by a series of cleanup discharges aimed at removing tritium from the torus. Measurements of tritium in the plasma and the exhaust were used to assess the efficiency of the various cleanup techniques. A preliminary estimate of the in-vessel tritium inventory is presented.
1. Introduction
2. Experimental arrangement
In November 1991, tritium was used in J E T to fuel tokamak plasmas [1]. This experiment differed from full D - T operation in two obvious ways: 1) the mixture of deuterium and tritium was relatively dilute in tritium, and 2) only two discharges were run to produce high D - T fusion power (> 1 MW). The tritium was introduced using neutral beam injection, so that while 2 • 1 0 22 T atoms were introduced into the NBI system for the whole experiment, only ~ 1 0 21 T atoms (2 • 1012 Bq) were injected into the toms. The tritium was introduced into the plasma as an 80 keV neutral beam because this was a more efficient means of fuelling the plasma than using cold gas. The beams were on for about 3.5 s in a total discharge duration of 25 s. Thus while the T concentration in thc plasma reached ~ 10%, the proportion of total gas input was only 1.3% and due to outgassing from the walls, the contribution in the exhaust was only ~ 0.3%. The discharges immediately following the high power discharges were devoted to the study of tritium release from the toms. The goals of this "cleanup" experiment were to 1) deplete the torus of residual tritium, 2) reduce the release of tritium exhaust from the vessel to levels suitable for atmospheric release ( < 6 • 10 is T atoms/day), 3) compare the efficiency of different cleanup techniques. This paper describes the cleanup experiment and reports on the extent to which the above goals were met. The focus of the present work is on the tritium release from the torus: detritiation of the neutral beam systems is reported elsewhere [2,3].
2.1. Vessel configuration The main plasma facing surfaces on JET are carbon and beryllium components (fig. 1). Two belt limiters run toroidally along the outer wall, with beryllium tiles on the upper limiter and carbon on the lower. A set of continuous beryllium X-point target plates run toroidally along the bottom of the vessel. A similar set, made of carbon, is located on the top. In addition, there are carbon tiles protecting octant bellows, rf antennas, and the inner wall. Evaporation of beryllium is performed periodically on all these surfaces as a wall conditioning procedure. The vessel temperature was 300~
2.2. Plasma configuration The tritium-fuelled discharges consisted of two phases: a start-up phase with the plasma on the lower belt limiter, followed by a single-null upper X-point phase (see fig. 1). It was predominantly in the second phase, during which the plasma contacted the carbon target plates, that tritium injection occurred. Although returning to a limiter phase is usual for terminating a discharge, these discharges were terminated in the X-point configuration in an effort to keep the areas of tritium contamination localized. This plasma configuration was also adopted for the cleanup discharges following tritium introduction. In addition these cleanup discharges included sweeping of thc strike zone across the target plates. This was done in an attempt to produce net erosion in the regions where net deposition of material had occurred during
0022-3115/92/$05.00 9 1992 - Elsevier Science Publishers B.V. All rights reserved
144
P. Andrew et a L / The tritium cleanup experiment in JET
tritium discharges. Otherwise, tritium in these regions would be buried by further deposition. (However, this is probably an unrealistic goal, since the geometry of the target plates means many areas are always net-deposition regions.)
To roughingpumps
t
2.3. Exhaust system The 200 m 3 torus volume is pumped by four turbomolecular pumps providing a total of ~ 9 m3/s (D 2) pumping speed, and up to ~ 5 m3/s by the NBI system [2]. The turbo pumps are paired into two pumping boxes (fig. 2). Total and partial pressure gauges are located just above the turbo pumps. The ducts connecting the pumping boxes to the torus each have a conductance about 10 times the pumping speed of the turbo pump pair. Therefore, after a discharge, the pressures in the pumping boxes are nearly equal to that in the torus. To handle tritium, the normal backing line (mechanical pumps exhausted to the atmosphere) was substituted by a closed system comprising a cryopump and uranium getter beds [4]. Just prior to each discharge, the backing line was isolated from the cryopump so that the exhaust evolving after the discharge would accumulate downstream of the turbo pumps, using the backing line as a buffer volume ~ 6 m 3. During this time the exhaust activity could be measured using an ionization chamber a n d / o r sampled by a removable sample bottle. The activity of the gas bottle samples
C a r b o n X-point Target Plates
~ ryopump
L He Dewar" "
Sample bottles
Fig. 2. Schematic of the exhaust system as used for the tritium experiment.
was measured on a separate system. Before the next discharge, the accumulated exhaust was pumped onto the cryopump. About 2 times each day, the cryopump was regenerated into a 0.345 m 3 volume, at which time the activity of the gas was measured by a second ionization chamber (see fig. 2).
2.4. Tritium monitoring
Collector Probe i t
xx
Be
f'x
Belt Limiter
Midplane
C Belt Limiter
/ Be X-point Target Plates
Fig. l. Section of the JET vessel. The magnetic surfaces shown are those for the X-point phase of the tritium-fuelled discharges.
No less than five different methods were used to measure the quantity of tritium released from the vessel during cleanup operation following the tritium discharges. The tritium concentration in the plasma was measured by applying a short burst of deuterium neutral beams and measuring 14 MeV neutrons using a threshold reaction in silicon [5]. The ratio of D - T to D - D neutrons gives the T / D ratio in the plasma and is insensitive to ion temperature. The amount of tritium being pumped out of the torus was measured using a mass spectrometer located in one of the pumping boxes. Mass numbers 2, 3, 4 and 5 as well as the total pressure (Penning gauge) were monitored for 1000 s after the start of the discharge. The sensitivity of this method was limited by a background of HD~- ions [2]. Downstream of the turbo pumps, measurements of the activity were made 560 s after the start of each discharge using one of the two ionization chambers
P. Andrew et al. / The tritium cleanup experiment in JET
a) Position of the probe in torus Probe
Pla"--.~ sma...
~Carbon
guide
Protection of \ wall protection "" "-.... Cross-section of probe showing samples analysed for tritium
Graphite shield
//shSamples Field line direction
Samples(~)
ample older (graphite)
Inconel samples ~
./~ 8mm slit Samples(~)
Fig. 3. Collector probe geometry.
145
distance from the midplane as the surface of the protection tiles at this radius. The slits were oriented in the radial direction so that only particles with a significant velocity component perpendicular to the toroidal direction could be collected on the inconel. In this configuration, the inconel sample directly behind the slits is intended to represent parts of the vessel wall not covered with graphite tiles. The remainder of the inconel samples are well protected from any direct impact and should therefore indicate tritium levels to be found on the shielded parts of the vessel. The end of the graphite shield is representative of graphite protection tiles in this part of the vessel. Four probe samples were analysed either by outgassing at high temperature, combining with oxygen and analysing for tritiated water (inconel samples), or by burning the entire sample in oxygen and analysing for tritiated water (carbon shield). Three samples were inconel sides of the collector, disposed around the collector as in fig. 3. The fourth sample was a piece of the graphite shield 34.9 • 7.4 mm 2 and 2.2 mm thick. This method meant that the probe was exposed to air prior to analysis. It has been observed [8] that such exposure tends to release tritium, and therefore the actual tritium concentration before atmospheric exposure may have been higher than that measured.
3. Cleanup procedure ( I C l i n fig. 2) down to ~ 3 x 1 0 5 T / D . Below that, the total amount of tritium collected per regeneration continued to be monitored even though the shot to shot release was no longer measured. When less than 6 • 1018 T atoms/day were being released, exhaust from the torus was routed to the atmosphere. The tritium release to the atmosphere was monitored by a detector in the exhaust stack [6]. When the exhaust was allowed to build up in the backing line it was also possible to sample the exhaust and analyse the gas sample on a remote system [7]. The virtue of this technique is its accuracy and much higher sensitivity. Tritium was detectable right up until the end of the experimental campaign (February 1992). 2.5. Collector probe
A collector probe was present in the JET vessel throughout the entire time that the closed exhaust system was used, after which it was removed for analysis of the tritium content. Thus an indication of residual in-vessel tritium could be obtained well in advance of the shutdown. The probe consisted of a graphite shield with two diametrically opposed 8 mm slits, and a ten-sided collector assembly, with each side consisting of three inconel samples each 12.7 x 9 mm 2 (fig. 3). The probe was mounted vertically in the torus (fig. 1) so that the bottom of the shield was at the same
Approximately 40 h after the last tritium fuelled discharge, a planned sequence of "cleanup" discharges started (see table 1). The first eight roughly identical ohmically heated discharges were aimed at revealing the shot to shot evolution of the tritium release. These were followed by a discharge of an alternative configuration: it started on the upper belt limiter, moved to the inner wall, and ended on the lower belt limiter. This was done to probe other surfaces for tritium. To compare the cleaning efficiency of helium versus deuterium plasmas, 3He was used to fuel a series of five discharges. After a brief resumption of the normal experimental program, two more days were devoted to detritiation of the vessel, because the tritium release level was still above the target value of ~ 6 x 10 is T atoms/day. Each day consisted of ~ 12 discharges in deuterium ending in high density disruptions, followed by a soak of the torus in deuterium gas at ~ 2 Pa pressure for a few hours. At this stage the regular experimental program was once again resumed. About two weeks after the initial main introduction of tritium, the level of tritium evolving from the torus was sufficiently low that the torus exhaust was routed to the atmosphere once again. At this stage, glow discharge cleaning of the torus was performed. GDC was not attempted earlier be-
146
P. Andrew et a L / The tritium cleanup experiment in JET
cause the closed exhaust system could not handle the high gas load.
ICI 10~
....... ~ ........ ~ T r i t i u m FueLled
9 9
]
Sample Bottle D l Soak
4. Results and discussion 10,, ~ I
~!~.~---I~isrupti
....
4. l. Exhaust measurements
The integrated tritium release for the 560 s period following the start of the plasmas is shown in fig. 4, as a function of the shot number. The two highest values correspond to the main tritium fuelled discharges. During the subsequent cleanup phase, the amount of tritium released per discharge decayed with shot number, N, roughly as e x p ( - N / N o) with N 0 = 10 for the first ~ 10 discharges and N o = 55 for the next ~ 100 discharges. Dual decay constants have been observed for H / D changeover experiments on T F T R [10] and have been modeled for isotopic exchange experiments in J E T [11]. The total amount of gas released in a 560 s period after the tritium fuelled discharges is about 50% of the total gas input, in agreement with previous gas accounting measurements [12]. The fractional tritium recovery for the tritium fuelled discharges is only about 12% for the same period. This agrees qualitatively with the recovery of T atoms after D - D discharges as
Table 1 Sequence of cleanup discharges Shot number Type
Comments
26150-26157 26159
D fuelling Ended in a disruption
Standard ohmic pulse Top belt, inner wall, lower belt 26160-26165 Standard ohmic pulse 26166-26170 Standard ohmic pulse 26171-26172 Standard ohmic pulse 26173 Top belt, inner wall, lower belt 26174-26178 Neutral beam heating 26180-26181 26182
Standard ohmic pulse Neutral beam heating
26184-26187
High density disruptions
26188-26189 26190-26243
Standard ohmic pulse Normal experimental program Additional heating and planned disruptions Soak torus in D 2 Additional heating and planned disruptions Soak torus in D 2 Normal experimental program
26245-26258
26260-26271
26272-
3He fuelling D fuelling No disruption 4-8 MW peak power 14 MW peak power Planned disruption
0-2 MW NBI 0-5 MW ICRH 0-2 MW NBI 0-11 MW ICRH
~
1
~.~i~ o o
Helium Plasmas l()/7
~
26140
.......
o
cm,9~ cz
o
~ 9
o[ .13
t , , , , , , , , , i ..... 26190 26240 Shot N u m b e r
, , 26290
Fig. 4. Tritium release after tokamak pulses as a function of the shot number. Dashed lines indicate decay constants of 10 and 55 discharges. Error bars shown are typical for RGA data. The error estimates for the ionization chamber and sample bottle data are typically _+30 and +_15% respectively.
compared with the number generated in the same discharge [12]. This could mean that atoms from the central plasma (i.e. injected T o or fusion product tritons) get more deeply trapped in the walls than particles in the boundary plasma. Alternatively the smaller T recovery may simply be due to dilution of the fuelled gas with the wall deuterium inventory. The first probing plasma, no. 26159, released more than two times as much tritium than the preceding discharge. However, this discharge also ended in an unplanned disruption. Disruptions are known to release an extra large quantity of gas following the discharge [9]. It is therefore not clear whether the enhanced tritium release was a result of the plasma visiting new surfaces or the disruption. The switch from D plasmas to 3He plasmas resulted in a ~ 2.5-fold reduction in the amount of tritium released per discharge. This suggests that surface recombination of T and D atoms, and subsequent release as D T molecules, is an important tritium release mechanism. It should be noted that there was a substantial amount of deuterium in these discharges. While ~ 3.5 • 1022 D atoms were needed to fuel a deuterium cleanup discharge, only ~ 8 • 10 21 D atoms (needed to ensure reliable breakdown) and ~ 2.5 • 1021 3He atoms were used to fuel the helium discharges. Much less gas in total is used as compared to the D-only discharges because of the smaller capacity of the walls for pumping helium. For the same reason, the plasma content is expected to be > 50% 3He, despite the larger contribution of D to the fuelling. This is supported by the observation of a fivefold decrease in D light (indicating a reduction in deuterium flux). A b o u t three times less deuterium appeared in the exhaust gas
147
P. Andrew et al. / The tritium cleanup experiment in JET [
10 :o
........ ~ II
lOJ')
~ ......... 9 N o v e m b e r 1991
~'l
9 '1 [
~
n --
560s l160s --
I() ~ ,< Target Outgassing Rate
9 1() I~
.........
26000
v/- - G D C ~, ('ollector Probe Removed
t,,,
......
r .........
26500 27000 Shot Number
27500
Fig. 5. As in fig. 4, but showing levels of tritium for sample bottles only, and including the periods before and after the tritium experiment. Samples for both 560 and 1160 s collection periods are shown.
Fig. 5 shows the release of tritium per discharge, but on a broader time scale. Only gas sample bottle results are shown since only this technique had sufficient sensitivity to measure tritium before and long after the tritium experiment. Some of the bottles sampled the exhaust at 560 s, while others sampled at 1160 s. The time dependence of the release rate [7] indicates that the longer collection time results in 1.4 _+ 0.2 times more tritium recovery. The data before the tritium experiment represent the quantity of T atoms released during a campaign of high fusion yield D - D discharges. This is followed by the first introduction of tritium in JET, first as a few discharges with trace quantities, and then as two discharges producing > 1 M W fusion power. No observable discontinuity in the shot to shot decay exists as a result of the glow discharge cleaning. After two months and more than 1000 tokamak discharges, the tritium release from the torus is almost down to the level which can be attributed to D - D fusion triton generation. 4.2. Plasma measurements
indicating that the T / D ratio for the helium discharges was slightly higher than for the deuterium discharges. Additional heating by deuterium neutral beams made very little difference to the quantity of tritium released. Disruptive discharges, on the other hand, resulted in approximately two times more tritium being released per discharge. The disruptions also resulted in larger amounts of deuterium release, the ratio for T / D being about the same as for discharges ending in "soft landings". Each deuterium soaking of the torus released more tritium than the discharges immediately preceding them (see fig. 4). However, if these releases are normalized by the quantity of deuterium used (a few tokamak discharges worth), the T / D ratio of the release is very nearly the same as for the tokamak discharges. It should be noted that although the soaking procedure consumed more time than discharges for equivalent tritium release, it is not known whether purges of shorter duration would have accomplished the same result. A 20 min period of G D C in deuterium released 5 x 1017 T atoms, i.e. about the same as four tokamak discharges at this time. However, an amount of gas equivalent to about 100 tokamak discharges was used, meaning relatively small T / D in the exhaust. Even though G D C attacks hidden surfaces and sputters material, the deuterium gas is not given a chance to equilibrate with the surfaces: deuterium flows through the torus in about 30 s and is not in the molecular flow regime. Therefore the deuterium gas may not have sufficient time to mix with gas in the walls. G D C in helium removed negligible tritium.
Neutron measurements showed that moving the plasma from one surface to another produced very little difference in T / D ratio of the plasma. This suggests that the tritium was not localized at the start of the cleanup. The value of T / D of the plasma from shot to shot indicated the same two time constant release behavior as the exhaust gas measurements [11]. 4.3. Surface concentrations
The results for the four collector probe samples are shown in table 2. These values can be used to make a crude estimate of the total vessel inventory when the probe was removed. If the tritium concentration of each sample is taken to be the same as the areas they were meant to represent (see table 2) the estimated total vessel inventory is 1.8 • 102o T atoms (3.3 x 1011 Bq). The uncertainty in this estimate is about a factor of 2 either way. This is about 17% of the total amount of tritium injected into the torus, and corresponds to a time ~ 250 discharges after the tritium introduction (see fig. 5). Forthcoming analysis of many in-vessel
Table 2 Tritium content of boundary probe samples Sample Material Surface Tritium number area release (• 10 4 m 2) (T/m 2) 1 2, 3 4
Inconel 3.46 Inconel 3.54, 3.49 Graphite 2.58
Corresponding toms area (m 2)
9.7:>< 10 t6 ~ 90 0.23-1.1 • 10 t++ ~ 250 1.5 • 10/8 ~ 110
148
P. Andrew et al. / The tritium cleanup experiment in JET
components for tritium [13] will give a more precise estimate of the tritium inventory. The vessel inventory estimate above is consistent with the tritium deficit between the total torus input and the sum of the exhaust recovery (as measured by the second ion chamber, IC2 in fig. 2) [2] for the same time: between 0-2.7 x 1020 T atoms. The total tritium release from the time of the probe removal to the end of operation is ~ 5 x 1019 T atoms.
5. Conclusions Two weeks (250 tokamak discharges) after the first introduction of tritium in the J E T torus (1021 T atoms), it was possible to reduce the in-vessel tritium inventory to a level between 0.5-2.7 x 1020 T atoms. This inventory estimate is supported by a preliminary surface concentration measurement. In the same period, the levels of tritium being released from the torus had been reduced to a level suitable for atmospheric release ( < 6 x 1018 T a t o m s / day). M e a s u r e m e n t of the T / D ratio in the plamsa was consistent with the exhaust, and indicated that the tritium in the vessel was not localized. Tritium release from a single discharge was (i) higher when the discharge ended in a disruption, (ii) lower when helium was substituted for deuterium as the fuelling gas. Additional heating made little difference. Soaking the torus in D e was found to be equally effective per unit exhaust gas as pulsing in the later stages of the cleanup. For all of the above techniques, the T / D ratio in the exhaust follows a common trend.
Acknowledgement One of us (P.A.) is grateful to the Canadian Fusion Fuels Technology Project for support.
References [1] JET Team, Nucl. Fusion 32 (1992) 187. [2] G. Saibene, R. Sartori, P. Andrew, J. How, Q. King and A.T. Peacock, to be published in Fusion Eng. Des. [3] H.D. Falter et al., these Proceedings (PSI-10), J. Nucl. Mater. 196-198 (1992) 1131. [4] J.L. Hemmerich et al., to be published in Fusion Eng. Des. [5] G. Sadler, O.N. Jarvis, P. van Belle and M. Pillon, Rev. Sci. Instr. 61 (1990) 3175. [6] C.J. Caldwell Nichols et al., to be published in Fusion Eng. Des. [7] D.H.J. Goodall, P. Andrew, J. Ehrenberg, G.M. McCracken, A.T. Peacock, G. Saibene and R. Sartori, these Proceedings, J. Nucl. Mater. 196-198 (1992) 1002. [8] J.P. Coad, A. Gibson, A.D. Haigh, G. Kaveney and J. Orchard, Proc. 18th EPS Conf. Berlin, 1991, Plasma Phys. Control. Fusion 15C, part III (1991) 81. [9] J. Ehrenberg et al., J. Nucl. Mater. 162-164 (1989) 63. [10] P.H. Lamarche, H.F. Dylla, P.J. McCarthy and M. UIrickson, J. Vac. Sci. Technol. 14 (1990) 1198. [11] L.D. Horton et al., these Proceedings (PSI-10), J. Nucl. Mater. 196-198 (1992) 139. [12] R. Sartori, G. Saibene, D.H.J. Goodall, E. Usselmann, P. Coad and D. Holland, J. Nucl. Mater. 176 & 177 (1990) 624. [13] A.T. Peacock, to be presented at the 17th Symp. on Fus. Tech., Rome, 1992.
Journal of Nuclear Materials 196-198 (1992) 143-148 North-Holland
The tritium cleanup experiment in JET P. Andrew, C.J. Caldwell-Nichols, J.P. Coad, K.J. Dietz, J. Ehrenberg, D.H.J. Goodall, J. Hemmerich, L. Horton, J. How, R. Lasser, P. Lomas, M. Loughlin, G.M. McCracken, A. Peacock, G. Saibene, R. Sartori, P. Thomas and T. Winkel JET Joint Undertaking, Abingdon, OxJordshire 0)(14 3EA, United Kingdom
The first tritium fuelled discharges in JET were followed by a series of cleanup discharges aimed at removing tritium from the torus. Measurements of tritium in the plasma and the exhaust were used to assess the efficiency of the various cleanup techniques. A preliminary estimate of the in-vessel tritium inventory is presented.
1. Introduction
2. Experimental arrangement
In November 1991, tritium was used in J E T to fuel tokamak plasmas [1]. This experiment differed from full D - T operation in two obvious ways: 1) the mixture of deuterium and tritium was relatively dilute in tritium, and 2) only two discharges were run to produce high D - T fusion power (> 1 MW). The tritium was introduced using neutral beam injection, so that while 2 • 1 0 22 T atoms were introduced into the NBI system for the whole experiment, only ~ 1 0 21 T atoms (2 • 1012 Bq) were injected into the toms. The tritium was introduced into the plasma as an 80 keV neutral beam because this was a more efficient means of fuelling the plasma than using cold gas. The beams were on for about 3.5 s in a total discharge duration of 25 s. Thus while the T concentration in thc plasma reached ~ 10%, the proportion of total gas input was only 1.3% and due to outgassing from the walls, the contribution in the exhaust was only ~ 0.3%. The discharges immediately following the high power discharges were devoted to the study of tritium release from the toms. The goals of this "cleanup" experiment were to 1) deplete the torus of residual tritium, 2) reduce the release of tritium exhaust from the vessel to levels suitable for atmospheric release ( < 6 • 10 is T atoms/day), 3) compare the efficiency of different cleanup techniques. This paper describes the cleanup experiment and reports on the extent to which the above goals were met. The focus of the present work is on the tritium release from the torus: detritiation of the neutral beam systems is reported elsewhere [2,3].
2.1. Vessel configuration The main plasma facing surfaces on JET are carbon and beryllium components (fig. 1). Two belt limiters run toroidally along the outer wall, with beryllium tiles on the upper limiter and carbon on the lower. A set of continuous beryllium X-point target plates run toroidally along the bottom of the vessel. A similar set, made of carbon, is located on the top. In addition, there are carbon tiles protecting octant bellows, rf antennas, and the inner wall. Evaporation of beryllium is performed periodically on all these surfaces as a wall conditioning procedure. The vessel temperature was 300~
2.2. Plasma configuration The tritium-fuelled discharges consisted of two phases: a start-up phase with the plasma on the lower belt limiter, followed by a single-null upper X-point phase (see fig. 1). It was predominantly in the second phase, during which the plasma contacted the carbon target plates, that tritium injection occurred. Although returning to a limiter phase is usual for terminating a discharge, these discharges were terminated in the X-point configuration in an effort to keep the areas of tritium contamination localized. This plasma configuration was also adopted for the cleanup discharges following tritium introduction. In addition these cleanup discharges included sweeping of thc strike zone across the target plates. This was done in an attempt to produce net erosion in the regions where net deposition of material had occurred during
0022-3115/92/$05.00 9 1992 - Elsevier Science Publishers B.V. All rights reserved
144
P. Andrew et a L / The tritium cleanup experiment in JET
tritium discharges. Otherwise, tritium in these regions would be buried by further deposition. (However, this is probably an unrealistic goal, since the geometry of the target plates means many areas are always net-deposition regions.)
To roughingpumps
t
2.3. Exhaust system The 200 m 3 torus volume is pumped by four turbomolecular pumps providing a total of ~ 9 m3/s (D 2) pumping speed, and up to ~ 5 m3/s by the NBI system [2]. The turbo pumps are paired into two pumping boxes (fig. 2). Total and partial pressure gauges are located just above the turbo pumps. The ducts connecting the pumping boxes to the torus each have a conductance about 10 times the pumping speed of the turbo pump pair. Therefore, after a discharge, the pressures in the pumping boxes are nearly equal to that in the torus. To handle tritium, the normal backing line (mechanical pumps exhausted to the atmosphere) was substituted by a closed system comprising a cryopump and uranium getter beds [4]. Just prior to each discharge, the backing line was isolated from the cryopump so that the exhaust evolving after the discharge would accumulate downstream of the turbo pumps, using the backing line as a buffer volume ~ 6 m 3. During this time the exhaust activity could be measured using an ionization chamber a n d / o r sampled by a removable sample bottle. The activity of the gas bottle samples
C a r b o n X-point Target Plates
~ ryopump
L He Dewar" "
Sample bottles
Fig. 2. Schematic of the exhaust system as used for the tritium experiment.
was measured on a separate system. Before the next discharge, the accumulated exhaust was pumped onto the cryopump. About 2 times each day, the cryopump was regenerated into a 0.345 m 3 volume, at which time the activity of the gas was measured by a second ionization chamber (see fig. 2).
2.4. Tritium monitoring
Collector Probe i t
xx
Be
f'x
Belt Limiter
Midplane
C Belt Limiter
/ Be X-point Target Plates
Fig. l. Section of the JET vessel. The magnetic surfaces shown are those for the X-point phase of the tritium-fuelled discharges.
No less than five different methods were used to measure the quantity of tritium released from the vessel during cleanup operation following the tritium discharges. The tritium concentration in the plasma was measured by applying a short burst of deuterium neutral beams and measuring 14 MeV neutrons using a threshold reaction in silicon [5]. The ratio of D - T to D - D neutrons gives the T / D ratio in the plasma and is insensitive to ion temperature. The amount of tritium being pumped out of the torus was measured using a mass spectrometer located in one of the pumping boxes. Mass numbers 2, 3, 4 and 5 as well as the total pressure (Penning gauge) were monitored for 1000 s after the start of the discharge. The sensitivity of this method was limited by a background of HD~- ions [2]. Downstream of the turbo pumps, measurements of the activity were made 560 s after the start of each discharge using one of the two ionization chambers
P. Andrew et al. / The tritium cleanup experiment in JET
a) Position of the probe in torus Probe
Pla"--.~ sma...
~Carbon
guide
Protection of \ wall protection "" "-.... Cross-section of probe showing samples analysed for tritium
Graphite shield
//shSamples Field line direction
Samples(~)
ample older (graphite)
Inconel samples ~
./~ 8mm slit Samples(~)
Fig. 3. Collector probe geometry.
145
distance from the midplane as the surface of the protection tiles at this radius. The slits were oriented in the radial direction so that only particles with a significant velocity component perpendicular to the toroidal direction could be collected on the inconel. In this configuration, the inconel sample directly behind the slits is intended to represent parts of the vessel wall not covered with graphite tiles. The remainder of the inconel samples are well protected from any direct impact and should therefore indicate tritium levels to be found on the shielded parts of the vessel. The end of the graphite shield is representative of graphite protection tiles in this part of the vessel. Four probe samples were analysed either by outgassing at high temperature, combining with oxygen and analysing for tritiated water (inconel samples), or by burning the entire sample in oxygen and analysing for tritiated water (carbon shield). Three samples were inconel sides of the collector, disposed around the collector as in fig. 3. The fourth sample was a piece of the graphite shield 34.9 • 7.4 mm 2 and 2.2 mm thick. This method meant that the probe was exposed to air prior to analysis. It has been observed [8] that such exposure tends to release tritium, and therefore the actual tritium concentration before atmospheric exposure may have been higher than that measured.
3. Cleanup procedure ( I C l i n fig. 2) down to ~ 3 x 1 0 5 T / D . Below that, the total amount of tritium collected per regeneration continued to be monitored even though the shot to shot release was no longer measured. When less than 6 • 1018 T atoms/day were being released, exhaust from the torus was routed to the atmosphere. The tritium release to the atmosphere was monitored by a detector in the exhaust stack [6]. When the exhaust was allowed to build up in the backing line it was also possible to sample the exhaust and analyse the gas sample on a remote system [7]. The virtue of this technique is its accuracy and much higher sensitivity. Tritium was detectable right up until the end of the experimental campaign (February 1992). 2.5. Collector probe
A collector probe was present in the JET vessel throughout the entire time that the closed exhaust system was used, after which it was removed for analysis of the tritium content. Thus an indication of residual in-vessel tritium could be obtained well in advance of the shutdown. The probe consisted of a graphite shield with two diametrically opposed 8 mm slits, and a ten-sided collector assembly, with each side consisting of three inconel samples each 12.7 x 9 mm 2 (fig. 3). The probe was mounted vertically in the torus (fig. 1) so that the bottom of the shield was at the same
Approximately 40 h after the last tritium fuelled discharge, a planned sequence of "cleanup" discharges started (see table 1). The first eight roughly identical ohmically heated discharges were aimed at revealing the shot to shot evolution of the tritium release. These were followed by a discharge of an alternative configuration: it started on the upper belt limiter, moved to the inner wall, and ended on the lower belt limiter. This was done to probe other surfaces for tritium. To compare the cleaning efficiency of helium versus deuterium plasmas, 3He was used to fuel a series of five discharges. After a brief resumption of the normal experimental program, two more days were devoted to detritiation of the vessel, because the tritium release level was still above the target value of ~ 6 x 10 is T atoms/day. Each day consisted of ~ 12 discharges in deuterium ending in high density disruptions, followed by a soak of the torus in deuterium gas at ~ 2 Pa pressure for a few hours. At this stage the regular experimental program was once again resumed. About two weeks after the initial main introduction of tritium, the level of tritium evolving from the torus was sufficiently low that the torus exhaust was routed to the atmosphere once again. At this stage, glow discharge cleaning of the torus was performed. GDC was not attempted earlier be-
146
P. Andrew et a L / The tritium cleanup experiment in JET
cause the closed exhaust system could not handle the high gas load.
ICI 10~
....... ~ ........ ~ T r i t i u m FueLled
9 9
]
Sample Bottle D l Soak
4. Results and discussion 10,, ~ I
~!~.~---I~isrupti
....
4. l. Exhaust measurements
The integrated tritium release for the 560 s period following the start of the plasmas is shown in fig. 4, as a function of the shot number. The two highest values correspond to the main tritium fuelled discharges. During the subsequent cleanup phase, the amount of tritium released per discharge decayed with shot number, N, roughly as e x p ( - N / N o) with N 0 = 10 for the first ~ 10 discharges and N o = 55 for the next ~ 100 discharges. Dual decay constants have been observed for H / D changeover experiments on T F T R [10] and have been modeled for isotopic exchange experiments in J E T [11]. The total amount of gas released in a 560 s period after the tritium fuelled discharges is about 50% of the total gas input, in agreement with previous gas accounting measurements [12]. The fractional tritium recovery for the tritium fuelled discharges is only about 12% for the same period. This agrees qualitatively with the recovery of T atoms after D - D discharges as
Table 1 Sequence of cleanup discharges Shot number Type
Comments
26150-26157 26159
D fuelling Ended in a disruption
Standard ohmic pulse Top belt, inner wall, lower belt 26160-26165 Standard ohmic pulse 26166-26170 Standard ohmic pulse 26171-26172 Standard ohmic pulse 26173 Top belt, inner wall, lower belt 26174-26178 Neutral beam heating 26180-26181 26182
Standard ohmic pulse Neutral beam heating
26184-26187
High density disruptions
26188-26189 26190-26243
Standard ohmic pulse Normal experimental program Additional heating and planned disruptions Soak torus in D 2 Additional heating and planned disruptions Soak torus in D 2 Normal experimental program
26245-26258
26260-26271
26272-
3He fuelling D fuelling No disruption 4-8 MW peak power 14 MW peak power Planned disruption
0-2 MW NBI 0-5 MW ICRH 0-2 MW NBI 0-11 MW ICRH
~
1
~.~i~ o o
Helium Plasmas l()/7
~
26140
.......
o
cm,9~ cz
o
~ 9
o[ .13
t , , , , , , , , , i ..... 26190 26240 Shot N u m b e r
, , 26290
Fig. 4. Tritium release after tokamak pulses as a function of the shot number. Dashed lines indicate decay constants of 10 and 55 discharges. Error bars shown are typical for RGA data. The error estimates for the ionization chamber and sample bottle data are typically _+30 and +_15% respectively.
compared with the number generated in the same discharge [12]. This could mean that atoms from the central plasma (i.e. injected T o or fusion product tritons) get more deeply trapped in the walls than particles in the boundary plasma. Alternatively the smaller T recovery may simply be due to dilution of the fuelled gas with the wall deuterium inventory. The first probing plasma, no. 26159, released more than two times as much tritium than the preceding discharge. However, this discharge also ended in an unplanned disruption. Disruptions are known to release an extra large quantity of gas following the discharge [9]. It is therefore not clear whether the enhanced tritium release was a result of the plasma visiting new surfaces or the disruption. The switch from D plasmas to 3He plasmas resulted in a ~ 2.5-fold reduction in the amount of tritium released per discharge. This suggests that surface recombination of T and D atoms, and subsequent release as D T molecules, is an important tritium release mechanism. It should be noted that there was a substantial amount of deuterium in these discharges. While ~ 3.5 • 1022 D atoms were needed to fuel a deuterium cleanup discharge, only ~ 8 • 10 21 D atoms (needed to ensure reliable breakdown) and ~ 2.5 • 1021 3He atoms were used to fuel the helium discharges. Much less gas in total is used as compared to the D-only discharges because of the smaller capacity of the walls for pumping helium. For the same reason, the plasma content is expected to be > 50% 3He, despite the larger contribution of D to the fuelling. This is supported by the observation of a fivefold decrease in D light (indicating a reduction in deuterium flux). A b o u t three times less deuterium appeared in the exhaust gas
147
P. Andrew et al. / The tritium cleanup experiment in JET [
10 :o
........ ~ II
lOJ')
~ ......... 9 N o v e m b e r 1991
~'l
9 '1 [
~
n --
560s l160s --
I() ~ ,< Target Outgassing Rate
9 1() I~
.........
26000
v/- - G D C ~, ('ollector Probe Removed
t,,,
......
r .........
26500 27000 Shot Number
27500
Fig. 5. As in fig. 4, but showing levels of tritium for sample bottles only, and including the periods before and after the tritium experiment. Samples for both 560 and 1160 s collection periods are shown.
Fig. 5 shows the release of tritium per discharge, but on a broader time scale. Only gas sample bottle results are shown since only this technique had sufficient sensitivity to measure tritium before and long after the tritium experiment. Some of the bottles sampled the exhaust at 560 s, while others sampled at 1160 s. The time dependence of the release rate [7] indicates that the longer collection time results in 1.4 _+ 0.2 times more tritium recovery. The data before the tritium experiment represent the quantity of T atoms released during a campaign of high fusion yield D - D discharges. This is followed by the first introduction of tritium in JET, first as a few discharges with trace quantities, and then as two discharges producing > 1 M W fusion power. No observable discontinuity in the shot to shot decay exists as a result of the glow discharge cleaning. After two months and more than 1000 tokamak discharges, the tritium release from the torus is almost down to the level which can be attributed to D - D fusion triton generation. 4.2. Plasma measurements
indicating that the T / D ratio for the helium discharges was slightly higher than for the deuterium discharges. Additional heating by deuterium neutral beams made very little difference to the quantity of tritium released. Disruptive discharges, on the other hand, resulted in approximately two times more tritium being released per discharge. The disruptions also resulted in larger amounts of deuterium release, the ratio for T / D being about the same as for discharges ending in "soft landings". Each deuterium soaking of the torus released more tritium than the discharges immediately preceding them (see fig. 4). However, if these releases are normalized by the quantity of deuterium used (a few tokamak discharges worth), the T / D ratio of the release is very nearly the same as for the tokamak discharges. It should be noted that although the soaking procedure consumed more time than discharges for equivalent tritium release, it is not known whether purges of shorter duration would have accomplished the same result. A 20 min period of G D C in deuterium released 5 x 1017 T atoms, i.e. about the same as four tokamak discharges at this time. However, an amount of gas equivalent to about 100 tokamak discharges was used, meaning relatively small T / D in the exhaust. Even though G D C attacks hidden surfaces and sputters material, the deuterium gas is not given a chance to equilibrate with the surfaces: deuterium flows through the torus in about 30 s and is not in the molecular flow regime. Therefore the deuterium gas may not have sufficient time to mix with gas in the walls. G D C in helium removed negligible tritium.
Neutron measurements showed that moving the plasma from one surface to another produced very little difference in T / D ratio of the plasma. This suggests that the tritium was not localized at the start of the cleanup. The value of T / D of the plasma from shot to shot indicated the same two time constant release behavior as the exhaust gas measurements [11]. 4.3. Surface concentrations
The results for the four collector probe samples are shown in table 2. These values can be used to make a crude estimate of the total vessel inventory when the probe was removed. If the tritium concentration of each sample is taken to be the same as the areas they were meant to represent (see table 2) the estimated total vessel inventory is 1.8 • 102o T atoms (3.3 x 1011 Bq). The uncertainty in this estimate is about a factor of 2 either way. This is about 17% of the total amount of tritium injected into the torus, and corresponds to a time ~ 250 discharges after the tritium introduction (see fig. 5). Forthcoming analysis of many in-vessel
Table 2 Tritium content of boundary probe samples Sample Material Surface Tritium number area release (• 10 4 m 2) (T/m 2) 1 2, 3 4
Inconel 3.46 Inconel 3.54, 3.49 Graphite 2.58
Corresponding toms area (m 2)
9.7:>< 10 t6 ~ 90 0.23-1.1 • 10 t++ ~ 250 1.5 • 10/8 ~ 110
148
P. Andrew et al. / The tritium cleanup experiment in JET
components for tritium [13] will give a more precise estimate of the tritium inventory. The vessel inventory estimate above is consistent with the tritium deficit between the total torus input and the sum of the exhaust recovery (as measured by the second ion chamber, IC2 in fig. 2) [2] for the same time: between 0-2.7 x 1020 T atoms. The total tritium release from the time of the probe removal to the end of operation is ~ 5 x 1019 T atoms.
5. Conclusions Two weeks (250 tokamak discharges) after the first introduction of tritium in the J E T torus (1021 T atoms), it was possible to reduce the in-vessel tritium inventory to a level between 0.5-2.7 x 1020 T atoms. This inventory estimate is supported by a preliminary surface concentration measurement. In the same period, the levels of tritium being released from the torus had been reduced to a level suitable for atmospheric release ( < 6 x 1018 T a t o m s / day). M e a s u r e m e n t of the T / D ratio in the plamsa was consistent with the exhaust, and indicated that the tritium in the vessel was not localized. Tritium release from a single discharge was (i) higher when the discharge ended in a disruption, (ii) lower when helium was substituted for deuterium as the fuelling gas. Additional heating made little difference. Soaking the torus in D e was found to be equally effective per unit exhaust gas as pulsing in the later stages of the cleanup. For all of the above techniques, the T / D ratio in the exhaust follows a common trend.
Acknowledgement One of us (P.A.) is grateful to the Canadian Fusion Fuels Technology Project for support.
References [1] JET Team, Nucl. Fusion 32 (1992) 187. [2] G. Saibene, R. Sartori, P. Andrew, J. How, Q. King and A.T. Peacock, to be published in Fusion Eng. Des. [3] H.D. Falter et al., these Proceedings (PSI-10), J. Nucl. Mater. 196-198 (1992) 1131. [4] J.L. Hemmerich et al., to be published in Fusion Eng. Des. [5] G. Sadler, O.N. Jarvis, P. van Belle and M. Pillon, Rev. Sci. Instr. 61 (1990) 3175. [6] C.J. Caldwell Nichols et al., to be published in Fusion Eng. Des. [7] D.H.J. Goodall, P. Andrew, J. Ehrenberg, G.M. McCracken, A.T. Peacock, G. Saibene and R. Sartori, these Proceedings, J. Nucl. Mater. 196-198 (1992) 1002. [8] J.P. Coad, A. Gibson, A.D. Haigh, G. Kaveney and J. Orchard, Proc. 18th EPS Conf. Berlin, 1991, Plasma Phys. Control. Fusion 15C, part III (1991) 81. [9] J. Ehrenberg et al., J. Nucl. Mater. 162-164 (1989) 63. [10] P.H. Lamarche, H.F. Dylla, P.J. McCarthy and M. UIrickson, J. Vac. Sci. Technol. 14 (1990) 1198. [11] L.D. Horton et al., these Proceedings (PSI-10), J. Nucl. Mater. 196-198 (1992) 139. [12] R. Sartori, G. Saibene, D.H.J. Goodall, E. Usselmann, P. Coad and D. Holland, J. Nucl. Mater. 176 & 177 (1990) 624. [13] A.T. Peacock, to be presented at the 17th Symp. on Fus. Tech., Rome, 1992.