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Release behavior of tritium from graphite heavily irradiated by neutrons
loo
Journal
of Nuclear Materials 99 ( 19X I ) I W- IO6 North-Holland Publishing Company
RELEASE BEHAVIOR OF TRITIUM FROM GRAPHITE HEAVILY IRRADIATED BY NEUTRONS Masakatsu
SAEKI
Division of Chemistry, Japan Atomic Energy Research Institute, Tokai, Ihurakr, Jupan Received
10 December
1980; in revised form
13 February
198 I
The release behavior of tritium formed in graphite has been investigated as a function of radiation damage by means of isochronal annealing of samples heavily irradiated by neutrons. The lithium impurities in graphite were estimated as the source of tritium formation. The main chemical form of released tritium was hydrogen accompanied by a small quantity of methane. No other hydrocarbons could be detected. Tritiated water was always measured, but the formation mechanism was experimentally confirmed as the secondary oxidation of released HT molecule. The release spectrum of tritium in isochronal annealing was shifted to a higher heating temperature with the increase of the neutron fluence received by the graphite crystal. A relationship was established between the amount of tritium released up to a certain temperature and the degree of graphitization of the sample.
1. Introduction
Graphite and carbon materials already have a lot of successful applications in nuclear technology. For a high temperature gas cooled reactor (HTGR), pyrolytic carbon has demonstrated its usefulness as an effective barrier against the release of fission products and tritium. In an HTGR, tritium is primarily produced by ternary fission due to thermal neutrons. Tritium is also generated by neutron induced nuclear reactions with impurities in the graphite [ 1,2]. In order to real&e zero release of radioactivity from nuclear plants in future, a much better understanding of tritium migration in graphites will be required. On the other hand, pyrolytic graphite is one of the candidate materials for the protection of the first wall from the attack of energetic plasma particles in a controlled thermonuclear reactor (CTR) [3]. The interaction of energetic plasma particles with the protection materials in a CTR will produce impurities for the plasma by processes such as sputtering and blistering. Thus, the reactions of atomic hydrogen isotopes with carbon materials have been studied by different authors using either hydrogen or deuterium atoms at thermal energies [4-61 or energetic ion beams of hydrogen isotopes [710]. However, the studies for reactions of energetic tritium with carbon materials are very limited and very little information is available about subsequent chemi0022-3115/81/0000-0000/$02.50
0 1981 North-Holland
cal interactions of tritium after its implantation into the graphite matrix [ll]. The fate of tritium after being adsorbed by carbon materials must be known to estimate the tritium inventory in a CTR. In the previous paper, the chemical reactions of energetic tritium atoms with powdered graphite have been studied by thermal neutron irradiation of closely packed graphite under an atmosphere of 3He. Evidence has been presented for the occurrence of hot chemical reactions between tritium and graphite with the formation of C-T bonds [12]. The main purpose of the present paper is to investigate the thermal release behavior of tritium from graphite as a function of radiation damage induced by neutron irradiation.
2. Experimental 2.1. Materials
Several kinds of graphites were used in this experiment. The samples mainly used was supplied by the Pechiney Company in France. The graphite, Pechiney Q,, has been used as neutron moderator in the power plant of Tokai-1, which is an advanced Calder Hall type reactor, in the Japan Atomic Power Company. Other samples are as follows; a natural graphite purified by the Ventron Corporation (Alfa Catalog stock no. 00642);
101
M. Sueki / Releuse of tritium from gruphite
an artificial graphite decomposed from a silicon carbide,which was prepared by Sasaki et al. [13]. All samples were primarily used after heat treatment in pure helium atmosphere for 10 hours at 1000°C or for one hour at 165O”C, but they were also used as received in order to compare the surface effect on the release behavior of tritium. Helium-3 was obtained from the Mound Laboratory with a stated purity of 99.8 mol%. 2.2. Sample preparation and irradiation Heavily irradiated graphite was spared from the monitors for the graphite of the Tokai-1 power plant through the Graphite Material Laboratory in JAERI. The monitors were prepared in order to measure the change on mechanical strength and thermal conductivity, and each size of specimens was 6.35 mm diam. X 76.2 mm. Thus, these specimens were irradiated in the Tokai -1 power plant. The irradiation temperature was 250400°C. The maximum neutron fluence of the sample was 1.66 X lo*’ n,,/cm* for thermal neutrons and 1.88 X lOI n,/cm’ for fast neutrons (E,( > 2 MeV). The reference samples were prepared as follows: The powdered graphites were packed closely into quartz ampoules, then these were evacuated. After being filled with 3He gas to a pressure of around 1 X lo* Pa, the ampoules were sealed. In order to investigate the effect of the distribution of tritium in graphite to the release behavior, LiNO, was mixed with graphite sample to approximately 50 mg LiNO, per one gram sample; in this case, 3He was not introduced. The samples were subsequently irradiated at the ambient temperature (about 40°C) in the T pipe of the JRR-4 reactor in JAERI for 5 min in order to implant tritium atoms into graphites. The thermal neutron flux was 8 X lOI n ,,,/cm’. s at the irradiation port. 2.3. Sample analysis Fig. 1 shows the apparatus for investigation of tritium release behavior from the irradiated sample and of the chemical forms of the released tritium. The heavily irradiated samples were usually used after being powdered by filing; the particle size was under 100 pm. An ahquot of the powdered sample or of the reference samples was usually weighed out in an alumina boat, then it was set in the alumina reaction tube. After removing the air in the reaction system by sweeping helium gas, the temperature was raised from room temperature to 1350°C at the rate of S”C/min and was kept constant for 30 min at the maximum temperature. In order to analyze the chemical form of
fr A
To
b-l A
4 ING
Fig. 1. Schematic diagram of experimental apparatus. (A) Steel bomb of He, (B) charcoal trap for purification at liquid N, temperature, (C) alumina boat and sample, (D) electric furnace, (E) P,O, column, (F) 4-way valve, (G) charcoal trap for released gases, (H) propylene carbonate separation column, (I) proportional counter, (J) flow rate meter.
tritium, the released tritiated gases were first trapped for a certain period on a charcoal bed at liquid nitrogen temperature and then injected into a separation column of propylene carbonate (6 mm I.D. X 5 m) by heating the trap up to 250°C. In order to determine the amount of tritium released as a function of the temperature, the gaseous tritium released upon heating was introduced directly into the proportional counter. The fraction of tritiated water, HTO, was measured by trapping in a water bubbler, when the amount was determined as a ‘function of temperature. The total amount of HTO was usually measured by trapping on a P,O, column and after dissolving the P,O, into water, the activity was counted by a liquid scintillation counter. Total amount of tritium incorporated into the irradiated sample was determined by burning a weighed aliquot in an oxygen atmosphere and the tritium released as HTO was trapped in a water bubbler, then the activity was determined by the liquid scintillation counter. 2.4. Characterization of graphite In the present experiment, the samples were characterized by measuring the lattice spacing C,,, with the X-ray diffractometer, Rigaku model CN 2028. The measurements were performed under following conditions: radiation: CuK,; slit system: divergence slit-l”, receiving slit-O.5 mm with Ni filter, scattering slit-l”.
with carbon atom of graphite [ 121. During the annealing process the bonds would be ruptured and the dissociated tritium atoms would diffuse to the surface, where they might collide with each other or with impurity hydrogen atom forming HT or T2. This process would be the most conceivable pathway for the main released form of hydrogen, HT. Tritiated water was another chemical form released from irradiated graphites. In the present experiment, all irradiated samples were exposed to the laboratory air at least once after the irradiation. The heavily irradiated samples were stored for several years only wrapped by a plastic sheet in a lead-made box. Thus, the sample surface was in an equilibrium with an atmospheric water. If tritium atoms incorporated in the graphite exchanged with hydrogen atoms of adsorbed water, the tritiated water would be released rather at low temperature [ 16,171. However, significant amount of HTO was not detected below 400°C. Table 1 shows the detected values of tritium as HTO in percent of total activity from samples which were variously stored until analysis. The results indicated that any differences in stored condition hardly affected the amounts of HTO released, and the values always stayed around 10%. When an alumina boat was preheated up to 135O”C, the observed amount of tritiated water was reduced to about one-half of its initial value. Furthermore, the value decreased to only 0.5% of total activity by using a graphite boat which was similarly preheated. In the next experiment, the release behavior of HTO was studied as a function of heating temperature. The results are shown in fig. 2 with the release curve for HT. From the figure, it is clearly understood that the fraction of HTO is directly proportional to the HT fraction. Thus, it seems conceivable at the present stage that the detected tritiated water was formed from the secondary reaction of released HT with molecular oxygen released from the proximity of the graphite sample in the reaction system.
3. Results and discussion 3.1. Formation graphite
reaction of tritium
in heuvi& irradiated
The total amounts of tritium remaining in the graphite irradiated up to 1.14 X IO*’ n,/cm* varied widely from sample to sample, the value ranged from 14.3 to 29.0 pCi/g graphite at the end of irradiation. This suggests that the main sources of tritium are the impurities in the graphite, because the distribution of the impurities in graphites is not usually uniform. The concentration of boron in Pechiney Q, was reported as 0.15 ppm [14]. This value will predict negligible amount of tritium according to the “B(nt,2a)T reaction, under the fast neutron flux of 2.38 X 10” n,/cm*.s, Ent > 2 MeV [ 151. The concentration of a lithium impurity was not reported about this graphite, and the value could not be easily measured, because it was too low. However, the concentration was assumed between lOO50 ppb from literature values reported for various graphites [ 1,2]. The activity of tritium from the ’ Li(n ,h, a)T reaction was calculated approximately IO pCi/g graphite using the minimum value for estimated lithium concentration and the average thermal neutron flux, ca. 10’” n ,,, /cm2. s [ 151. This results indicated that almost all tritium detected in the sample was produced according to the 6Li(n th, a)T reaction. 3.2. Chemical forms of released tritium The chemical forms of released tritium were identified by radio gas-chromatography. The main chemical form was hydrogen, HT, and a small quantity of methane, CH,T, was also detected. No other hydrocarbons could be detected in the present experiment, while tritiated water, HTO, was always accompanied as much as around 10%. As reported previously, the tritium incorporated into graphite combines chemically
Table I Amounts
of HTO released
from Pechiney
Q, irradiated
under various
conditions
and stored variously
until analysis
Pretreatment temperature
Neutron fluence (n/cm’ )
Stored conditions
HTO (%)
1000°C. IO h Room temperature Room temperature 368°C” 248°C ‘)
2.40x 2.40x 2.40X 4.80X 1.66X
O-5 days in air 1.3X1&‘Pa,45h 1.3X 10e2 Pa, 28O”C, 42 h ca. 6 years in air ca. 7 years in air
9.45 1.8 Il.5 9.6 12.5k2.6 9.5 k 2.5
‘) Temperature
of the sample during
IO’” 10’0 lOI IO*” 102’
irradiation
M. Saeki / Release
of tntium
from
103
gruphite
3.3. Influence of radiation damage on thermal release behavior of tritium
$0 Fig. 2. Release heating time.
320
280
240
As mentioned above, the released chemical form of tritium incorporated into graphite is primarily hydrogen, HT. Thus, the gases released from irradiated samples were directly introduced into the proportional counter after passing through a P,O, column in order to compare the release spectra of HT in isochronal annealing from variously irradiated graphites. One set of the results are shown in fig. 3, which were obtained by using the graphite of Pechiney Q,. The introducing method of tritium into graphite will be described in detail in appendix. In the figure, curve A was obtained from the powdered Pechiney Q, sample into which tritium was implanted by means of method- 1 described in the Appendix. The release consisted of two broad peaks. The first started around 500°C and the second began at around 750°C. Curve B was from the sample which was first irradiated in the power plant up to 4.8 x 102” nth/cm 2 (method-2) then after being powdered by filing, tritium was introduced again into the powder by means of method- 1. It is simply because the amount of tritium produced by the first way in the sample was too small for this kind of experiment. In this case, the first peak of curve A disappeared, and the second peak starting around 750°C tailed to a higher temperature. Curve C was obtained by using the powdered sample irradiated in the power plant up to 1.66 X 10” ne,/cm2 (method-2). The peak starting around 750°C became a minor peak and the third peak began at around 1000°C. Since it could be experimentally proven as mentioned in appendix that the distribution of tritium in the graphite particle had minor importance for the release behavior of tritium from the sample, the observed changings in the release spectra shown in fig. 3 should be caused by the increase of radiation damage with increasing neutron fluence. Similar experiments were performed using a well graphitized graphite which was formed by decomposing a silicon carbide [ 131 and also
HEATING TIME (min)
behavior
of HT
and
HTO
as a function
of
Table 2 shows the total amounts of CH,T released from different graphites. The pretreatment effectively reduced the released amount, while the value did not vary among the graphites. Boothe and Ache recently reported the chemical reaction of recoil tritium with graphite [ 111. They studied the chemical form of tritium released from graphite upon heating, and found a lot of tritiated hydrocarbons containing methane, all C, hydrocarbons and propane. The maximum yield of hydrocarbons was beyond 15%. However, they used a batch method, that is, after irradiation the sealed ampoules were heated at elevated temperatures for a certain period, for example, at 1000°C for 4 h. Many authors reported the evolution of hydrocarbons from carbon materials at lower temperatures below 1000°C [l&20], while the composition of hydrocarbons evolved has not been well established. Thus, the hydrocarbons released from the graphite could decompose thermally into radicals, such as, methyl, ethyl and so on. These resulted radicals would react with released HT from the irradiated sample and would change into tritiated hydrocarbons in the sealed ampoule. These successive reactions may explain the discrepancy between the present and Boothe’s results.
Table 2 Total amounts
of CH,T
Pretreatment temperature
released
from various
1650°C
in percent
of total tritium
Sample Pechiney
Room temperature 1000°C
graphites
Q,
0.43 Ik 0.05 0.21 kO.02 _
Sic decomposed
Natural
0.45 ro.05 0.32 1tO.03
0.40c0.04 _
_
0.29 k 0.03
- 1350 -1250
1
100
I
I
150
200 TIME
bin)
in iaochronal
annealing
HEATING Fig. 3. Release spectra
of tritium
I
300
a natural graphite. The introduction of tritium into these graphites was carried out by means of method - 1. The release spectra of tritium from both graphites were shifted to a lower temperature and the heating up to 1200°C allowed almost complete release of tritium from the samples. The remaining amounts of tritium in the samples were measured after annealing experiments. The results are summarized in table 3 with the neutron fluences received by the samples and the lattice spacing C,,,. The amount of tritium retained after annealing experiments in samples increase with winding the space of basal plane. From these experimental results, it seems qualitatively that some relationships exist between the release behavior of tritium from the graphite sample and the
of tritium
in graphite
after annealing
Kinds of graphite
Neutron fluence (n/cm’)
Natural SC decomposed Pechiney Q, Pechiney Q , (R15-17) Pechiney Q, (303-47)
2.40X lOI 2.40X 10lh 2.40x 10lh 4. x0 k I 0 2’1 1.66X 10:’
experiments
irradiated
Retention
Pechiney
Q, graphite
degree of order in the graphite crystal. As shown in table 3, the lattice spacing was obviously widened with increasing neutron fluence. This phenomenon was well studied by many authors [21]. It has been established that the c-spacing expands and the u-spacing contracts linearly with received dose of irradiation, and this is due to the reduction of the graphitization in the irradiated sample. A number of formulas were presented in order to describe the degree of graphitization by the c-spacing [22-251. In the present case, the following equation proposed by Ishikawa et al. [24] was used to express the degree of the order in the graphite crystal: Degree of graphitization
(W)
= (0.3440 - d/O.3440 - 0.3354) x 100, where d = Co,,,/2
in units of percent
(“r)
1
350
from variously
using
Table 3 Retention
I
I
250
in unit of nm. The logarithms
of total tritium Cl,,,,, (nm)
Degree of graphitization (%Fc)
0.3+0.1 0.2kO.I 0.5 CO.2 3.0t0.9 IZ.Xt3.7
0.6709 0.671 I 0.673 I 0.6742
99.4 9x.3 86.6 80.2
0.676 I
69.2
(1) of the
M. Saeki / Release of tritium from gruphite
105
4. Conclusions
RETENTION
70 DEGREE
80 OF
90
GRAPHITIZATION
100 (X)
Fig. 4. Relationship between the degree of graphitization and the amounts of tritium released up to certain temperatures and
the reciprocals of percent retention after annealing experiments.
integral yields of tritium released up to 800, 1000 and 1200°C are plotted versus the calculated degree of graphitization in fig. 4, while the fraction released as HTO was corrected as it is directly proportional to the HT fraction based on the reason mentioned above. In the figure, the logarithms of the reciprocals of the residual tritium in the samples after annealing experiments (table 3) are also plotted versus the calculated values. Although linearity of the lines is not sufficient, it can be concluded that the release behavior of tritium relates directly to the degree of order in the sample crystal. By the experiment of isothermal anneal using the similarly irradiated graphite, it has been confirmed that a correlation exists between the diffusion coefficient of tritium in graphite and the degree of graphitization [26]. Thus, the relationship shown in fig. 4 might be explained by the changes in the diffusibility of tritium in graphite crystal rather than by those in the trapping characteristics in damaged graphite.
The release behavior of tritium formed in graphite has been investigated as a function of radiation damage by means of isochronal annealing of the samples heavily irradiated by neutrons in the Tokai-1 power plant. It was estimated that tritium found in the samples was mainly formed from lithium impurities in graphite by the 6Li(n,,,a)T reaction. The main chemical form of released tritium was hydrogen, HT, accompanied by a small quantity of methane, CH,T. No other hydrocarbons could be detected in the present experiment. However, tritiated water, HTO, was always detected, and the formation mechanism was experimentally confirmed as the secondary oxidation of released HT molecule from graphite. The portions of each chemical form were ca. 90% for HT, ca. 10% for HTO and ca. 0.5% for CH;T, respectively. The release spectrum of tritium in isochronal annealing shifted to higher heating temperatures with the increase of the neutron fluence received by the graphite sample. A relationship was established between the amount of tritium released up to a certain temperature and the degree of graphitization of the sample.
Appendix
For this experiment, three different ways were adopted in order to implant tritium into graphite samples. Method- 1 was that the powdered graphite was irradiated with 3He and the tritium from the 3He(n,,,p)T reaction was implanted into the sample. The advantage of this method is that the surface of the graphite can be kept clean. Method-2 was that the graphite was irradiated for a long time in the power plant, as the results considerable amount of tritium was produced probably from lithium impurities as mentioned in section 3.1. The distribution of tritium in the specimen might be homogeneous. Method-3 was that the powdered graphite well mixed with LiNO, was irradiated and the tritium from the 6Li(n,,,a)T reaction was introduced into the graphite. In this case, the LiNO, source for tritium has to be removed by washing with water, before the sample is used. Thus, this method was only employed in order to prove that the distribution of tritium in the sample did not give rise to different release behavior of tritium in isochronal annealing. The range of tritium in graphite has been reported as 1.5 pm for the 3He(n,,,p)T reaction and 33 pm for the
“Li(n,,,,a)T reaction, respectively [27]. In method-3. each particle of graphite was coated with a thin film of LiNO,, because the lithium salt was added to powdered graphite as aqueous solution, then the water was dried up. The tritium recoiled from a target with infinitesimal thickness has almost the same kinetic energy and is emitted isotropically, thus, a uniform concentration of tritium might be assumed in the region between the surface and the end of the range [28]. In the present case, since the maximum radius of the graphite powder was 50 pm, the region where tritium was almost homogeneously distributed occupied more than 95% of the whole volume of a particle of graphite. Thus, the distribution of tritium in graphite by means of method-2 can be approximated by the implantation of tritium into sample by means of method-3. In order to study the effects of the different distribution of tritium in graphite, an isochronal annealing was carried out with the sample prepared by method-3. The results obtained from this sample was very close to curve A shown in fig. 3, while the first peak starting around 500°C was slightly higher than that of the curve A. Since the sample was prepared under similar conditions (the kind of graphite, the period of irradiating time and approximate number of tritium atoms implanted into graphite) except the source of tritium, it should be conclude that the distribution of tritium in graphite hardly affected to the release behavior of tritium. At the present stage, a possible explanation is only that the distribution of tritium in graphite sample prepared by, method-l might be more homogeneous than that considered from the range of tritium from the ‘He(n,,,p)T reaction, because the source of ‘He can penetrate into open porosities of graphite.
Acknowledgements The author would like to thank Mr. Tamotsu Saito for the measurement of X-ray diffractometry on unirradiated samples. He greatly appreciates the helpful comments of Dr. Enzo Tachikawa.
References [II B.W. Gainey. GA-A 13461 (1976). 121 E.L. Compere, S.H. Freid and C.W. Nestor, ORNL-TM[31 [41 [51 [61 171 PI [91 [101
[I41 [I51 (161 1171
4303 ( 1974). R.W. Conn, J. Nucl. Mater. 76/77 (197X) 103. B.J. Wood and H. Wise, J. Phys. Chem. 73 ( 1969) 134X. R.K. Gould, J. Chem. Phys. 63 (1975) 1825. M. Balooch and D.R. Olander, J. Chem. Phys. 63 (I 975) 4772. R. Behrisch, J. Bohdansky, G.H. Oetjen, J. Roth, G. Schilling and H. Verbeek, J. Nucl. Mater. 60 (1976) 32 I. J. Roth, J. Bohdansky, W. Poschenrieder and M.K. Sinha. J. Nucl. Mater. 63 (1976) 222. R.A. Langley, R.S. Blewer and J. Roth, J. Nucl. Mater. 76,‘77 (1978) 313. G. Staudenmaier, J. Roth, R. Behrisch, J. Bohdansky. W. Eckstein. P. Staib and M. Matteson, J. Nucl. Mater. 84 (1979) 149. T.E. Boothe and H.J. Ache, J. Phys. Chem. 82 (1978) 1362. M. Saeki and F.S. Rowland, to be published. S. Komori, Y. Sasaki and Y. Shima, J. Chem. Sot. Japan, Industrial Chem. Section 63 (1960) 37 (in Japanese). S. Sato, JAERI-memo 1380 (1964). T. Nogaito, private communication. G.C.R. Semiann, ORNL-2767 (1959) p. 190. K. Yamabe and H. Takahashi, Tanso (Carbon) IO2 ( 1980) 106 (in Japanese).
[I81 J.P. Redmond and P.L. Walker, Jr., Nature I86 (1960) 72. [I91 S. Yajima, S. Ichiba and M. Handa, J. At. Energy Sot. Japan 3 ( 1961) 942 (in Japanese). PO1 H. Imai. S. Nomura, T. Kurosawa, K. Fujii and Y. Sasaki, Carbon I6 (1978) 404. Radiation Damage in Graphite (PerPII J.H.W. Simmons, gamon, Oxford, 1965) p, 121. R.E. Franklin, Acta Cryst. 4 (1951) 253. PI ~231 G.E. Bacon, Acta Cryst. 4 (1951) 558. 1241 T. Ishikawa and S. Yoshizawa, J. Chem. Sot. Japan. Industrial Chem. Section 66 (1963) 14. 1251 J. Meting and J. Maire, Les Carbons I (I 965) 129. PI M. Saeki, to be published. Stopping Powers and Ranges in All Elements, v71 Hydrogen Eds. H.H. Anderson and J.F. Ziegler (Pergmon, Oxford, 1977). WI G.D. Cola and Hj. Matzke, EUR 3488e (1967).
Journal
of Nuclear Materials 99 ( 19X I ) I W- IO6 North-Holland Publishing Company
RELEASE BEHAVIOR OF TRITIUM FROM GRAPHITE HEAVILY IRRADIATED BY NEUTRONS Masakatsu
SAEKI
Division of Chemistry, Japan Atomic Energy Research Institute, Tokai, Ihurakr, Jupan Received
10 December
1980; in revised form
13 February
198 I
The release behavior of tritium formed in graphite has been investigated as a function of radiation damage by means of isochronal annealing of samples heavily irradiated by neutrons. The lithium impurities in graphite were estimated as the source of tritium formation. The main chemical form of released tritium was hydrogen accompanied by a small quantity of methane. No other hydrocarbons could be detected. Tritiated water was always measured, but the formation mechanism was experimentally confirmed as the secondary oxidation of released HT molecule. The release spectrum of tritium in isochronal annealing was shifted to a higher heating temperature with the increase of the neutron fluence received by the graphite crystal. A relationship was established between the amount of tritium released up to a certain temperature and the degree of graphitization of the sample.
1. Introduction
Graphite and carbon materials already have a lot of successful applications in nuclear technology. For a high temperature gas cooled reactor (HTGR), pyrolytic carbon has demonstrated its usefulness as an effective barrier against the release of fission products and tritium. In an HTGR, tritium is primarily produced by ternary fission due to thermal neutrons. Tritium is also generated by neutron induced nuclear reactions with impurities in the graphite [ 1,2]. In order to real&e zero release of radioactivity from nuclear plants in future, a much better understanding of tritium migration in graphites will be required. On the other hand, pyrolytic graphite is one of the candidate materials for the protection of the first wall from the attack of energetic plasma particles in a controlled thermonuclear reactor (CTR) [3]. The interaction of energetic plasma particles with the protection materials in a CTR will produce impurities for the plasma by processes such as sputtering and blistering. Thus, the reactions of atomic hydrogen isotopes with carbon materials have been studied by different authors using either hydrogen or deuterium atoms at thermal energies [4-61 or energetic ion beams of hydrogen isotopes [710]. However, the studies for reactions of energetic tritium with carbon materials are very limited and very little information is available about subsequent chemi0022-3115/81/0000-0000/$02.50
0 1981 North-Holland
cal interactions of tritium after its implantation into the graphite matrix [ll]. The fate of tritium after being adsorbed by carbon materials must be known to estimate the tritium inventory in a CTR. In the previous paper, the chemical reactions of energetic tritium atoms with powdered graphite have been studied by thermal neutron irradiation of closely packed graphite under an atmosphere of 3He. Evidence has been presented for the occurrence of hot chemical reactions between tritium and graphite with the formation of C-T bonds [12]. The main purpose of the present paper is to investigate the thermal release behavior of tritium from graphite as a function of radiation damage induced by neutron irradiation.
2. Experimental 2.1. Materials
Several kinds of graphites were used in this experiment. The samples mainly used was supplied by the Pechiney Company in France. The graphite, Pechiney Q,, has been used as neutron moderator in the power plant of Tokai-1, which is an advanced Calder Hall type reactor, in the Japan Atomic Power Company. Other samples are as follows; a natural graphite purified by the Ventron Corporation (Alfa Catalog stock no. 00642);
101
M. Sueki / Releuse of tritium from gruphite
an artificial graphite decomposed from a silicon carbide,which was prepared by Sasaki et al. [13]. All samples were primarily used after heat treatment in pure helium atmosphere for 10 hours at 1000°C or for one hour at 165O”C, but they were also used as received in order to compare the surface effect on the release behavior of tritium. Helium-3 was obtained from the Mound Laboratory with a stated purity of 99.8 mol%. 2.2. Sample preparation and irradiation Heavily irradiated graphite was spared from the monitors for the graphite of the Tokai-1 power plant through the Graphite Material Laboratory in JAERI. The monitors were prepared in order to measure the change on mechanical strength and thermal conductivity, and each size of specimens was 6.35 mm diam. X 76.2 mm. Thus, these specimens were irradiated in the Tokai -1 power plant. The irradiation temperature was 250400°C. The maximum neutron fluence of the sample was 1.66 X lo*’ n,,/cm* for thermal neutrons and 1.88 X lOI n,/cm’ for fast neutrons (E,( > 2 MeV). The reference samples were prepared as follows: The powdered graphites were packed closely into quartz ampoules, then these were evacuated. After being filled with 3He gas to a pressure of around 1 X lo* Pa, the ampoules were sealed. In order to investigate the effect of the distribution of tritium in graphite to the release behavior, LiNO, was mixed with graphite sample to approximately 50 mg LiNO, per one gram sample; in this case, 3He was not introduced. The samples were subsequently irradiated at the ambient temperature (about 40°C) in the T pipe of the JRR-4 reactor in JAERI for 5 min in order to implant tritium atoms into graphites. The thermal neutron flux was 8 X lOI n ,,,/cm’. s at the irradiation port. 2.3. Sample analysis Fig. 1 shows the apparatus for investigation of tritium release behavior from the irradiated sample and of the chemical forms of the released tritium. The heavily irradiated samples were usually used after being powdered by filing; the particle size was under 100 pm. An ahquot of the powdered sample or of the reference samples was usually weighed out in an alumina boat, then it was set in the alumina reaction tube. After removing the air in the reaction system by sweeping helium gas, the temperature was raised from room temperature to 1350°C at the rate of S”C/min and was kept constant for 30 min at the maximum temperature. In order to analyze the chemical form of
fr A
To
b-l A
4 ING
Fig. 1. Schematic diagram of experimental apparatus. (A) Steel bomb of He, (B) charcoal trap for purification at liquid N, temperature, (C) alumina boat and sample, (D) electric furnace, (E) P,O, column, (F) 4-way valve, (G) charcoal trap for released gases, (H) propylene carbonate separation column, (I) proportional counter, (J) flow rate meter.
tritium, the released tritiated gases were first trapped for a certain period on a charcoal bed at liquid nitrogen temperature and then injected into a separation column of propylene carbonate (6 mm I.D. X 5 m) by heating the trap up to 250°C. In order to determine the amount of tritium released as a function of the temperature, the gaseous tritium released upon heating was introduced directly into the proportional counter. The fraction of tritiated water, HTO, was measured by trapping in a water bubbler, when the amount was determined as a ‘function of temperature. The total amount of HTO was usually measured by trapping on a P,O, column and after dissolving the P,O, into water, the activity was counted by a liquid scintillation counter. Total amount of tritium incorporated into the irradiated sample was determined by burning a weighed aliquot in an oxygen atmosphere and the tritium released as HTO was trapped in a water bubbler, then the activity was determined by the liquid scintillation counter. 2.4. Characterization of graphite In the present experiment, the samples were characterized by measuring the lattice spacing C,,, with the X-ray diffractometer, Rigaku model CN 2028. The measurements were performed under following conditions: radiation: CuK,; slit system: divergence slit-l”, receiving slit-O.5 mm with Ni filter, scattering slit-l”.
with carbon atom of graphite [ 121. During the annealing process the bonds would be ruptured and the dissociated tritium atoms would diffuse to the surface, where they might collide with each other or with impurity hydrogen atom forming HT or T2. This process would be the most conceivable pathway for the main released form of hydrogen, HT. Tritiated water was another chemical form released from irradiated graphites. In the present experiment, all irradiated samples were exposed to the laboratory air at least once after the irradiation. The heavily irradiated samples were stored for several years only wrapped by a plastic sheet in a lead-made box. Thus, the sample surface was in an equilibrium with an atmospheric water. If tritium atoms incorporated in the graphite exchanged with hydrogen atoms of adsorbed water, the tritiated water would be released rather at low temperature [ 16,171. However, significant amount of HTO was not detected below 400°C. Table 1 shows the detected values of tritium as HTO in percent of total activity from samples which were variously stored until analysis. The results indicated that any differences in stored condition hardly affected the amounts of HTO released, and the values always stayed around 10%. When an alumina boat was preheated up to 135O”C, the observed amount of tritiated water was reduced to about one-half of its initial value. Furthermore, the value decreased to only 0.5% of total activity by using a graphite boat which was similarly preheated. In the next experiment, the release behavior of HTO was studied as a function of heating temperature. The results are shown in fig. 2 with the release curve for HT. From the figure, it is clearly understood that the fraction of HTO is directly proportional to the HT fraction. Thus, it seems conceivable at the present stage that the detected tritiated water was formed from the secondary reaction of released HT with molecular oxygen released from the proximity of the graphite sample in the reaction system.
3. Results and discussion 3.1. Formation graphite
reaction of tritium
in heuvi& irradiated
The total amounts of tritium remaining in the graphite irradiated up to 1.14 X IO*’ n,/cm* varied widely from sample to sample, the value ranged from 14.3 to 29.0 pCi/g graphite at the end of irradiation. This suggests that the main sources of tritium are the impurities in the graphite, because the distribution of the impurities in graphites is not usually uniform. The concentration of boron in Pechiney Q, was reported as 0.15 ppm [14]. This value will predict negligible amount of tritium according to the “B(nt,2a)T reaction, under the fast neutron flux of 2.38 X 10” n,/cm*.s, Ent > 2 MeV [ 151. The concentration of a lithium impurity was not reported about this graphite, and the value could not be easily measured, because it was too low. However, the concentration was assumed between lOO50 ppb from literature values reported for various graphites [ 1,2]. The activity of tritium from the ’ Li(n ,h, a)T reaction was calculated approximately IO pCi/g graphite using the minimum value for estimated lithium concentration and the average thermal neutron flux, ca. 10’” n ,,, /cm2. s [ 151. This results indicated that almost all tritium detected in the sample was produced according to the 6Li(n th, a)T reaction. 3.2. Chemical forms of released tritium The chemical forms of released tritium were identified by radio gas-chromatography. The main chemical form was hydrogen, HT, and a small quantity of methane, CH,T, was also detected. No other hydrocarbons could be detected in the present experiment, while tritiated water, HTO, was always accompanied as much as around 10%. As reported previously, the tritium incorporated into graphite combines chemically
Table I Amounts
of HTO released
from Pechiney
Q, irradiated
under various
conditions
and stored variously
until analysis
Pretreatment temperature
Neutron fluence (n/cm’ )
Stored conditions
HTO (%)
1000°C. IO h Room temperature Room temperature 368°C” 248°C ‘)
2.40x 2.40x 2.40X 4.80X 1.66X
O-5 days in air 1.3X1&‘Pa,45h 1.3X 10e2 Pa, 28O”C, 42 h ca. 6 years in air ca. 7 years in air
9.45 1.8 Il.5 9.6 12.5k2.6 9.5 k 2.5
‘) Temperature
of the sample during
IO’” 10’0 lOI IO*” 102’
irradiation
M. Saeki / Release
of tntium
from
103
gruphite
3.3. Influence of radiation damage on thermal release behavior of tritium
$0 Fig. 2. Release heating time.
320
280
240
As mentioned above, the released chemical form of tritium incorporated into graphite is primarily hydrogen, HT. Thus, the gases released from irradiated samples were directly introduced into the proportional counter after passing through a P,O, column in order to compare the release spectra of HT in isochronal annealing from variously irradiated graphites. One set of the results are shown in fig. 3, which were obtained by using the graphite of Pechiney Q,. The introducing method of tritium into graphite will be described in detail in appendix. In the figure, curve A was obtained from the powdered Pechiney Q, sample into which tritium was implanted by means of method- 1 described in the Appendix. The release consisted of two broad peaks. The first started around 500°C and the second began at around 750°C. Curve B was from the sample which was first irradiated in the power plant up to 4.8 x 102” nth/cm 2 (method-2) then after being powdered by filing, tritium was introduced again into the powder by means of method- 1. It is simply because the amount of tritium produced by the first way in the sample was too small for this kind of experiment. In this case, the first peak of curve A disappeared, and the second peak starting around 750°C tailed to a higher temperature. Curve C was obtained by using the powdered sample irradiated in the power plant up to 1.66 X 10” ne,/cm2 (method-2). The peak starting around 750°C became a minor peak and the third peak began at around 1000°C. Since it could be experimentally proven as mentioned in appendix that the distribution of tritium in the graphite particle had minor importance for the release behavior of tritium from the sample, the observed changings in the release spectra shown in fig. 3 should be caused by the increase of radiation damage with increasing neutron fluence. Similar experiments were performed using a well graphitized graphite which was formed by decomposing a silicon carbide [ 131 and also
HEATING TIME (min)
behavior
of HT
and
HTO
as a function
of
Table 2 shows the total amounts of CH,T released from different graphites. The pretreatment effectively reduced the released amount, while the value did not vary among the graphites. Boothe and Ache recently reported the chemical reaction of recoil tritium with graphite [ 111. They studied the chemical form of tritium released from graphite upon heating, and found a lot of tritiated hydrocarbons containing methane, all C, hydrocarbons and propane. The maximum yield of hydrocarbons was beyond 15%. However, they used a batch method, that is, after irradiation the sealed ampoules were heated at elevated temperatures for a certain period, for example, at 1000°C for 4 h. Many authors reported the evolution of hydrocarbons from carbon materials at lower temperatures below 1000°C [l&20], while the composition of hydrocarbons evolved has not been well established. Thus, the hydrocarbons released from the graphite could decompose thermally into radicals, such as, methyl, ethyl and so on. These resulted radicals would react with released HT from the irradiated sample and would change into tritiated hydrocarbons in the sealed ampoule. These successive reactions may explain the discrepancy between the present and Boothe’s results.
Table 2 Total amounts
of CH,T
Pretreatment temperature
released
from various
1650°C
in percent
of total tritium
Sample Pechiney
Room temperature 1000°C
graphites
Q,
0.43 Ik 0.05 0.21 kO.02 _
Sic decomposed
Natural
0.45 ro.05 0.32 1tO.03
0.40c0.04 _
_
0.29 k 0.03
- 1350 -1250
1
100
I
I
150
200 TIME
bin)
in iaochronal
annealing
HEATING Fig. 3. Release spectra
of tritium
I
300
a natural graphite. The introduction of tritium into these graphites was carried out by means of method - 1. The release spectra of tritium from both graphites were shifted to a lower temperature and the heating up to 1200°C allowed almost complete release of tritium from the samples. The remaining amounts of tritium in the samples were measured after annealing experiments. The results are summarized in table 3 with the neutron fluences received by the samples and the lattice spacing C,,,. The amount of tritium retained after annealing experiments in samples increase with winding the space of basal plane. From these experimental results, it seems qualitatively that some relationships exist between the release behavior of tritium from the graphite sample and the
of tritium
in graphite
after annealing
Kinds of graphite
Neutron fluence (n/cm’)
Natural SC decomposed Pechiney Q, Pechiney Q , (R15-17) Pechiney Q, (303-47)
2.40X lOI 2.40X 10lh 2.40x 10lh 4. x0 k I 0 2’1 1.66X 10:’
experiments
irradiated
Retention
Pechiney
Q, graphite
degree of order in the graphite crystal. As shown in table 3, the lattice spacing was obviously widened with increasing neutron fluence. This phenomenon was well studied by many authors [21]. It has been established that the c-spacing expands and the u-spacing contracts linearly with received dose of irradiation, and this is due to the reduction of the graphitization in the irradiated sample. A number of formulas were presented in order to describe the degree of graphitization by the c-spacing [22-251. In the present case, the following equation proposed by Ishikawa et al. [24] was used to express the degree of the order in the graphite crystal: Degree of graphitization
(W)
= (0.3440 - d/O.3440 - 0.3354) x 100, where d = Co,,,/2
in units of percent
(“r)
1
350
from variously
using
Table 3 Retention
I
I
250
in unit of nm. The logarithms
of total tritium Cl,,,,, (nm)
Degree of graphitization (%Fc)
0.3+0.1 0.2kO.I 0.5 CO.2 3.0t0.9 IZ.Xt3.7
0.6709 0.671 I 0.673 I 0.6742
99.4 9x.3 86.6 80.2
0.676 I
69.2
(1) of the
M. Saeki / Release of tritium from gruphite
105
4. Conclusions
RETENTION
70 DEGREE
80 OF
90
GRAPHITIZATION
100 (X)
Fig. 4. Relationship between the degree of graphitization and the amounts of tritium released up to certain temperatures and
the reciprocals of percent retention after annealing experiments.
integral yields of tritium released up to 800, 1000 and 1200°C are plotted versus the calculated degree of graphitization in fig. 4, while the fraction released as HTO was corrected as it is directly proportional to the HT fraction based on the reason mentioned above. In the figure, the logarithms of the reciprocals of the residual tritium in the samples after annealing experiments (table 3) are also plotted versus the calculated values. Although linearity of the lines is not sufficient, it can be concluded that the release behavior of tritium relates directly to the degree of order in the sample crystal. By the experiment of isothermal anneal using the similarly irradiated graphite, it has been confirmed that a correlation exists between the diffusion coefficient of tritium in graphite and the degree of graphitization [26]. Thus, the relationship shown in fig. 4 might be explained by the changes in the diffusibility of tritium in graphite crystal rather than by those in the trapping characteristics in damaged graphite.
The release behavior of tritium formed in graphite has been investigated as a function of radiation damage by means of isochronal annealing of the samples heavily irradiated by neutrons in the Tokai-1 power plant. It was estimated that tritium found in the samples was mainly formed from lithium impurities in graphite by the 6Li(n,,,a)T reaction. The main chemical form of released tritium was hydrogen, HT, accompanied by a small quantity of methane, CH,T. No other hydrocarbons could be detected in the present experiment. However, tritiated water, HTO, was always detected, and the formation mechanism was experimentally confirmed as the secondary oxidation of released HT molecule from graphite. The portions of each chemical form were ca. 90% for HT, ca. 10% for HTO and ca. 0.5% for CH;T, respectively. The release spectrum of tritium in isochronal annealing shifted to higher heating temperatures with the increase of the neutron fluence received by the graphite sample. A relationship was established between the amount of tritium released up to a certain temperature and the degree of graphitization of the sample.
Appendix
For this experiment, three different ways were adopted in order to implant tritium into graphite samples. Method- 1 was that the powdered graphite was irradiated with 3He and the tritium from the 3He(n,,,p)T reaction was implanted into the sample. The advantage of this method is that the surface of the graphite can be kept clean. Method-2 was that the graphite was irradiated for a long time in the power plant, as the results considerable amount of tritium was produced probably from lithium impurities as mentioned in section 3.1. The distribution of tritium in the specimen might be homogeneous. Method-3 was that the powdered graphite well mixed with LiNO, was irradiated and the tritium from the 6Li(n,,,a)T reaction was introduced into the graphite. In this case, the LiNO, source for tritium has to be removed by washing with water, before the sample is used. Thus, this method was only employed in order to prove that the distribution of tritium in the sample did not give rise to different release behavior of tritium in isochronal annealing. The range of tritium in graphite has been reported as 1.5 pm for the 3He(n,,,p)T reaction and 33 pm for the
“Li(n,,,,a)T reaction, respectively [27]. In method-3. each particle of graphite was coated with a thin film of LiNO,, because the lithium salt was added to powdered graphite as aqueous solution, then the water was dried up. The tritium recoiled from a target with infinitesimal thickness has almost the same kinetic energy and is emitted isotropically, thus, a uniform concentration of tritium might be assumed in the region between the surface and the end of the range [28]. In the present case, since the maximum radius of the graphite powder was 50 pm, the region where tritium was almost homogeneously distributed occupied more than 95% of the whole volume of a particle of graphite. Thus, the distribution of tritium in graphite by means of method-2 can be approximated by the implantation of tritium into sample by means of method-3. In order to study the effects of the different distribution of tritium in graphite, an isochronal annealing was carried out with the sample prepared by method-3. The results obtained from this sample was very close to curve A shown in fig. 3, while the first peak starting around 500°C was slightly higher than that of the curve A. Since the sample was prepared under similar conditions (the kind of graphite, the period of irradiating time and approximate number of tritium atoms implanted into graphite) except the source of tritium, it should be conclude that the distribution of tritium in graphite hardly affected to the release behavior of tritium. At the present stage, a possible explanation is only that the distribution of tritium in graphite sample prepared by, method-l might be more homogeneous than that considered from the range of tritium from the ‘He(n,,,p)T reaction, because the source of ‘He can penetrate into open porosities of graphite.
Acknowledgements The author would like to thank Mr. Tamotsu Saito for the measurement of X-ray diffractometry on unirradiated samples. He greatly appreciates the helpful comments of Dr. Enzo Tachikawa.
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