performance of ceramic breeder materials in the sibelius experiment

ELSEVIER Journal of Nuclear Materials 219 (1995) 259-264

lnm I

Performance of ceramic breeder materials in the SIBELIUS experiment

J.P. Kopasz a, C.E. Johnson a, D.L. Baldwin b a Argonne National Laboratory, Argonne, 1L 60439, USA

b Pacific Northwest Laboratories, Richland, WA 99352, USA

Abstract

Lithium-containing ceramics are among the leading candidates for use as tritium breeding materials in a fusion reactor. An issue affecting both the safety and economics of the reactor is the tritium inventory. The SIBELIUS experiment was designed to examine material compatibility between the different components of a breeder blanket (ceramic breeder, beryllium neutron multiplier, steel structural material) and to examine the tritium inventory in the ceramic and beryllium. The tritium inventories in four candidate ceramics, as determined by measurements at the end of life, were found to be low and unchanged by the presence of beryllium. The inventory increased in the order lithium zirconate < lithium oxide < lithium orthosilicate < lithium aluminate, with the inventory in the zirconate less than 0.03 wppm at 550°C.

1. Introduction

Lithium-containing ceramics are among the leading candidates for use as tritium-breeding materials in a fusion reactor. Breeder blanket designs utilizing lithium ceramics also employ a beryllium neutron multiplier to ensure a tritium-breeding ratio greater than one. For these designs to be acceptable from an economic and safety perspective, the tritium inventory in the ceramic and in the beryllium should be low. The SIBELIUS experiment is an E U / U S A collaborative test designed to assess the compatibility of beryl l ium/l i thium ceramics and beryl l ium/steel compacts in a neutron environ- ment. The test also provided the opportunity to assess the tritium inventory in several candidate ceramic breeder materials and in beryllium and to determine whether any interactions between the ceramics and beryllium affect the tritium inventory in these materials.

This paper focuses on end-of-life tests to determine the tritium inventory in the ceramic samples. End-of- life inventory measurements are less ambiguous than inventories determined from in-pile temperature change tests and provide crucial tritium inventory data for designers and modelers. However, sample charac-

0022-3115/95/$09.50 © 1995 Elsevier Science B.V. All rights SSDI 0022-3115(94)00400-5

teristics such as grain size may be affected by changes in the test conditions, such as temperature, or purge gas composition, which may complicate the interpreta- tion of the end-of-life inventories. The SIBELIUS data set is unique in that the samples were subjected to a minimal number of temperature changes and operated under constant purge-gas composition.

2. Experimental

The irradiation vehicle consisted of eight capsules, seven of which were independently purged with a H e - 0.1% H 2 gas mixture. The eighth capsule was not purged and contained beryllium pellets for lifetime and void swelling tests. One capsule contained Li4SiO 4 pebbles in contact with beryllium disks, and four others contained alternating disks of beryllium, steel, and ceramic in the order beryllium, steel, beryllium, ceramic. Each of the four capsules contained a different ceramic (either lithium orthosilicate, lithium aluminate, lithium oxide, or lithium zirconate). Two capsules contained only ceramic (Li20 and LiAIO2), for comparison of tritium release and inventory with the samples in contact with beryllium. The design of the irradi-

reserved

260 J.P. Kopasz et al. /Journal of Nuclear Materials 219 (1995) 259-264

ation capsules and the experimental conditions have been reported previously [1,2]. A summary of the ceramic properties is given in Table 1.

The specimens were irradiated for 1690 h with a neutron flux at mid-plane of 1.1 × 1014 n c m-2 s -1 (thermal) and 1.0 x 1014 n cm -2 s -1 (fast, > 1 MeV). The irradiation temperature was controlled and main- tained at 550°C for most of the experiment. One set of temperature transients was performed near the end of the irradiation period, with the temperature for the aluminate decreased to 500°C then returned to 550°C, and the silicate and zirconate decreased to 450°C then increased back to 550°C, while the temperature for the oxide sample remained constant. The temperature at the end of the experiment was 550°C.

After irradiation, the ceramic specimens were removed from the stack and examined. All ceramic specimens were intact. Tritium analysis was carried out on sections of the samples by high temperature anneals. Details of the experimental apparatus for the annealing have been described previously [3,4]. Each SIBELIUS sample disk was weighed and then split into approximate halves within an enclosed container. A weighed half-disk was used for each anneal test. The purge gas of He-0.1% H 2 was purified through a commercial dryer and oxygen-removal column rated for < 50 ppb total H 2 0 and 0 2. The purge gas flow rate was 100 cma/min. The all-quartz apparatus design allowed the weighed sample to be dropped into the heated zone after the system was stable at temperature. The anneal temperatures and times are listed in Table 2.

The tritium released during the anneals was not passed through any reducing column but was swept directly into a calibrated ionization chamber connected to a Keithly Model 617 Electrometer. All data acquisi- tion and control of the electrometer were performed by an on-line personal computer. The tritium was then passed through an oxidizing column of CuO at 500°C and collected in a pair of sequential water collection bubblers and measured by liquid scintillation counting. The sample counts were converted to Bq of tritium by comparison with a tritium standard directly traceable to National Institute of Standards and Technology. Prior to reaching the CuO column, the purge gas stream had ambient air added at about 10 cm3/min to provide oxygen in order to maintain the CuO in a

Table 1 Sample characteristics

Material

LizZrO 3 Li20 Li4SiO 4 LiAlO 2

Grain size (p,m) 1 ~ 20 ~ 20 0.4 Density (%TD) 70 80 91 75

Table 2 Anneal conditions

Sample Anneal Anneal % of total temperatures times released at each (°C) temperature

LiA10 2 550 ~ 24 h 42.4 650 ~ 24 h 52.0 750 ~ 24 h 5.0 850 ~ 48 h 0.6

Li2ZrO 3 550 ~ 24 h 84.9 850 ~ 24 h 15.1

Li20 850 only ~ 17 h 100

Li4SiO 4 850 only ~ 3 days 100

regenerated state. Liquid scintillation counting provided quantitative assay results, while the ionization chamber data provided real-time release curves. The collection bubblers were changed and measured once for each temperature step, providing the single quantitative data point for each anneal.

The inventory in some of the ceramic samples was determined by dissolution followed by scintillation counting. The dissolutions were performed after re- moving surface contamination due to steel activation products from the samples. Gamma spectroscopy re- vealed the presence of 6°C0, 54Mn, and 65Zn on all of the ceramic samples, indicating contamination from steel activation products. In addition, the gamma spectroscopy indicated a small amount of activity due to nsSb and 152Eu, 154Eu, and 155Eu in the oxide sample.

The zirconate sample contained a large number of activated species, the largest contributors being 95Zr and 9SNb. Since these activated species would not be removed by any surface treatment and would interfere with the tritium determination by scintillation counting, no dissolution was attempted for the zirconate samples.

The samples used for the dissolution tests were broken in pieces, and each piece placed in a tared glass vial and weighed. The lithium oxide samples were dissolved in 5 ml of distilled water in closed screw-cap vials to prevent escape of HTO vapor. The solutions were then transferred to 25 ml volumetric flasks and diluted to the appropriate volume. Aliquots were taken for liquid scintillation counting. Unlike the lithium oxide samples, the lithium aluminate and lithium silicate samples would not dissolve completely. Partial dissolutions were accomplished by adding 10 ml of an acid solution (approximately 6N HNO 3) to a screw-cap vial containing the sample, closing the vial, and heating for about 6 h. The acid solution was then decanted from the solid into a 100 ml volumetric flask, and a second 10 ml of acid solution added. The procedure was repeated, and the solid remaining was washed with

J.P. Kopasz et al. /Journal of Nuclear Materials 219 (1995) 259-264 261

10 ml of distilled water. The solid and vial were then dried and weighed to determine the amount of solid which had dissolved.

Dissolution of the oxide in distilled water allows for titration to determine the extent of hydrolysis of the sample prior to the dissolution. For that purpose, 2 ml aliquots of the diluted solution were pipetted into 50 ml Erlenmeyer flasks and titrated with 0.049 M HC1. The endpoint was determined using Bromocresol Pur- ple as an indicator. During the dissolution there is the possibility that some HT may form. If this occurs, calculations indicate that, because of the equilibrium among HT, H2, HTO, and H20, only a small fraction (<< 1%) of the tritium will remain in the form of HT. Therefore, this procedure does not take any special precautions to collect tritium in the form of HT gas formed during the dissolution, such as oxidation followed by collection of the HTO formed. In addition, estimates of the HTO vapor pressure above the solution under the conditions of the dissolution (closed vial) are expected to lead to errors less than 1%.

3. Results

3.1. Tritium inventories in ceramics

Results of the tritium anneals performed on the lithium aluminate sample are shown in Fig. 1. The final

35000

30000

25000

g 20oo0 6

15000

E 10000

5000

-5000

Tritium Release from LiAIO 2

8 5 0 ° C

7 5 0 ° C ~

650"C

i

550"C :

0 20 40 60 80

Time, h

90O

850

8O0

750

700 o.

650

E 600

550

500

450

4O0

100

Tritium release - - Temperature ....

Fig. I. Successive isothermal anneals for the lithium aluminate sample.

Table 3 Tritium inventories in SIBELIUS ceramics

Sample Inventory (MBq/g)

Dissolution Anneals Average

LiAIO 2 2029 +203 1659 _+116 1844 _+185 Li4SiO 4 874 ± 87 947 ± 66 911 ± 37 Li20 34.8± 3.5 36.5_+ 2.6 35.5_+ 0.7 Li2ZrO 3 ND 7.8± 0.6 ND

anneal was performed at 850°C for 48 h. During the second 24-h period of this anneal, an additional 0.1% of the tritium was released. For the lithium zirconate, one anneal was performed for 24 h at 550°C, followed by a second 24 h anneal at 850°C. The majority of the tritium (84.9%) was released during the anneal at 550°C. For lithium oxide only one anneal at 850°C was performed. No additional release was observed after 17 h. For the lithium silicate, the sample was annealed at 850°C. After 48 h tritium was still being released, so the anneal was continued. After 72 h no additional release was observed.

The tritium inventories determined from the dissolutions and the anneals are tabulated in Table 3. The lithium oxide results in the table have been corrected to give the inventory prior to any hydrolysis of the sample. The amount of hydrolysis was determined by titration of the lithium oxide sample solutions. For lithium oxide, two moles of lithium hydroxide would be formed for each mole of lithium oxide upon dissolution, giving 0.067 moles hydroxide/g of sample. If the sample was already hydrolyzed to lithium hydroxide, there would be 0.042 moles hydroxide/g of sample. Titrations of the solution formed upon dissolution indicated a hydroxide concentration of 0.041 mol /g of sample. This agrees with the concentration of hydroxide in lithium hydroxide of 0.042 m o l / g indicating that the lithium oxide samples had fully hydrolyzed prior to the analysis. Most likely this hydrolysis occurred during the handling and storage of the samples and not during the irradiation. Assuming the same amount of hydrolysis occurred in each sample, correcting the tritium inventory to t r i t ium/g of lithium oxide increases the inventory from 21.7 to 34.8 M B q / g for the dissolution tests and from 22.6 to 36.2 M B q / g for the anneal samples. The ternary ceramics are much less hygroscopic than lithium oxide and are expected to have undergone only slight hydrolysis during handling and storage. In addition, if hydrolysis occurred, the per- centage of weight gained due to hydrolysis is much less than that for lithium oxide, so errors in determining the tritium per unit weight of sample are reduced. It is estimated that the error in the inventories due to any hydrolysis of the ternary ceramics is less than 10%.

262 J.P. Kopasz et al. /Journal of Nuclear Materials 219 (1995) 259-264

3.2. Overall performance

The beryllium/ceramic system appears stable at 550°C for the irradiation fluence investigated. The lithium ceramics survived the irradiation intact without fractures. The ceramic disks were easily separated from the adjacent beryllium disks, with no macroscopic evi- dence for interactions between the ceramic and beryllium. Examinations with scanning electron microscopy indicated no changes in the ceramic microstructures at the beryl l ium/ceramic interface; however, some roughening of the beryllium surfaces was noted [2,5]. The stability of the beryllium/ceramic couple is proba- bly due to a protective coating of BeO, which appears to be unaffected by irradiation damage at the fluence investigated. As indicated in Table 3, the tritium inventory increased in the order Li2ZrO 3 < Li20 < LiaSiO 4 < LiAlO 2. These inventories were all within the range of values observed for other in-pile tests where beryllium was absent, suggesting that the beryllium has little or no effect on the tritium retention in the ceramic.

have an effect on the inventory, since tritium partial pressure increases in moving from the purge gas inlet (tritium partial pressure of zero) to the purge gas outlet (nonzero tritium partial pressure). This effect has been proposed as the reason for the increase in tritium inventory at the purge inlet compared to the purge outlet observed for the BEATRIX thin-ring samples [6]. A similar gradient was observed in the CRITIC-1 experiment [7]. However, for the SIBELIUS experiment the samples analyzed by Dienst et al. [5] were located by the purge gas inlet and would be expected to have lower inventories than those analyzed in this work, which were positioned by the purge gas outlet, if solubility was a main factor in determining inventory. Also, the differences between the inventories we determined and those determined by Dienst et al. cannot be attributed to reaction of the hygroscopic lithium oxide samples with water, as our values were corrected for this factor and any hydrolysis experienced by Dienst's samples would increase the discrepancy.

4. Discussion

4.1. Comparison of analysis methods

The tritium inventory in lithium oxide was determined by two methods (high temperature anneals and dissolution) at two different laboratories (Pacific Northwest Laboratories and Argonne National Labo- ratory). The results are in excellent agreement, with differences of less than 5%. The results indicate that all of the tritium was released during the high temperature anneals and that losses due to tritium in the gas phase (either as HT or HTO) for the dissolution method are insignificant. For lithium silicate, the agreement between the two methods is also very good. The great- est difference was found in the inventories determined for the aluminate samples, with a difference of about 20%. This difference is most likely due to the small amount of aluminate sample that dissolved (6 mg) and the inherent errors involved in the partial dissolution. No dissolution analysis was done for the zirconate sample.

4.2. Comparison with other SIBELIUS determinations

The inventories that we determined for the oxide (35.5 MBq/g) and the silicate (910 MBq/g) are lower than those determined by Dienst et al. [5] for SIBELIUS samples in the same capsules of 70.5 and 1219 MBq/g. The differences in the inventories for the SIBELIUS samples do not appear to be attributable to any differences in the samples or irradiation environ- ment. One may expect the position in the capsule to

Table 4 Comparison of tritium inventories

Ceramic Experiment Temper- Grain size Inventory ature (txm BOL) (MBq/g) (°C)

Li20 SIBELIUS 550 20 35.5 This work EXOTIC-2 610 7.5 18 BEATRIX-II 532-649 25-40 22 Solid BEATRIX-II 630 5.5 81-215 Ring SIBELIUS 550 20 70.5 Dienst CRITIC-1 600 55 106-353 VOM-15H 750 2 176-353

SIBELIUS 550 0.4 1844 This work TRIO 650 0.1 16.7 EXOTIC-6 495 0.4 148 (CEA) EXOTIC-6 485 1-10 2405 (ENEA)

SIBELIUS 550 20 910 This work SIBELIUS 550 20 1219 Dienst EXOTIC-6 360 10-30 1776

SIBELIUS 550 1 7.84 This work EXOTIC-6 450 1 2.96 (CEA) EXOTIC-6 455 2 7.03 (ENEA)

LiAIO 2

Li4SiO 4

LizZrO 3

J.P. Kopasz et al. /Journal of Nuclear Materials 219 (1995) 259-264 263

4.3. Comparisons with other in-pile tests

Table 4 compares the tritium inventories determined in the SIBELIUS experiment with those determined in other experiments. The tritium inventory that we determined for the SIBELIUS lithium oxide sample, 35.5 M B q / g (0.096 wppm), agrees with that determined in BEATRIX-II , 0.06 wppm, for the portion of the solid (temperature gradient) lithium oxide specimen in temperature regions calculated to be from 532 to 649°C [6]. It also agrees with the value for EXOTIC-2 lithium oxide of 0.05 wppm (temperature 610°C) [8]. The agreement is exceptional if one uses the SIBELIUS inventory prior to correction for hydrolysis of 22.2 M B q / g (0.064 wppm). This raises the possibility that hydrolysis may have also occurred for the BEATRIX-II temperature gradient and EXOTIC-2 samples. The SIBELIUS inventory values that we determined are about a factor of two to six lower than the inventories reported for the BEATRIX-II thin-ring specimen of 0.23-0.61 wppm [6], even though the average temperature in the thin-ring at shutdown was 630°C. The value obtained by Dienst et al. (0.20 wppm) [5] is comparable to the value reported for the BEATRIX-II thin ring sample positioned at the bottom near the purge inlet (0.23 wppm) [6]. The differences in tritium inventories determined for these experiments do not correlate with differences in sample grain size. The reasons for the variations in inventories are not clear at this time. One point which is clear is that the presence of hydrogen in the purge reduces the tritium inventory. The tritium inventories discussed above are all less than those determined in CRITIC-I , 0.3 to 1 ppm at 600°C [7], and VOM-15H, 0.5-1.0 ppm at 740°C [9], which were obtained with the samples under purge gases with lower hydrogen concentrations.

Comparisons of inventories for lithium aluminate samples indicate an even larger variation. For the SIBELIUS lithium aluminate sample, we determined a tritium inventory of 1844 M B q / g (5.2 wppm). In the TRIO experiment an inventory o f 16.7 M B q / g was found for LiAIO 2 (temperature, 650°C; grain size, 0.2 ixm) [10], while in EXOTIC-6, inventories ranged from 148 M B q / g (final temperature, 495°C) for the CEA- prepared aluminate to 2405 M B q / g (final temperature 485°C) for ENEA-prepared material. Kwast et al. [11] suggest the difference in inventory for the EXOTIC aluminates is related to differences in grain size (CEA aluminate grain size, 0.4 i~m; ENEA aluminate grain size, 1-10 ixm). However, the SIBELIUS sample, al- though at slightly higher temperature, has a grain size comparable to that of the CEA aluminate in EXOTIC, but an inventory comparable to the ENEA aluminate. This suggests that another variable is responsible for the difference in tritium inventories for aluminate sam- pies. The SIBELIUS tritium generation rate is about

70% higher than in EXOTIC. Thus, for a desorption release mechanism the inventories would be expected to be proportionately higher. However, this difference in tritium generation rates does not account for the more than order of magnitude difference in the inventories of similarly sized samples.

For the lithium orthosilicate, the tritium inventories determined at the end of life are in good agreement; however, these values do not agree with those determined from temperature-transient tests. We determined an inventory of 910 M B q / g for the SIBELIUS lithium orthosilicate pellet. Dienst et al. [5] determined a slightly higher inventory of 1219 M B q / g for a second SIBELIUS lithium orthosilicate pellet, and Kwast et al. [11] reported an inventory of 1776 M B q / g at the end-of-life for lithium orthosilicate at 360°C in EX- OTIC-6. These values are higher than expected based on calculations from in-pile tritium release transients after temperature changes [12]. In SIBELIUS the temperature transients indicate a tritium inventory of 442 M B q / g at 550°C (residence time of 1 h) [12]. This suggests that tritium inventories determined from temperature-transient tests may substantially underpredict tritium inventories.

A tritium inventory of 7.84 M B q / g was determined for the SIBELIUS lithium zirconate from annealing studies. This is slightly greater than thatt determined by Kwast et al. [11] of 2.96 M B q / g for CEA zirconate (1 txm grain diameter) but is in good agreement with that determined for larger grain-sized material (7.03 M B q / g for ENEA zirconate with 2 ~m grain diameter). The SIBELIUS zirconate was similar to the CEA zirconate used in EXOTIC; however, the inventory is more comparable to that for the larger ENEA zirconate. Again, the reason for these variations is un- known.

5. Conclusions

Because reliable tritium inventory data are so few, we can not draw firm conclusions at this time. Nonetheless, it appears that intimate contact between beryllium and lithium ceramics considered for tritium breeding materials has no detrimental effects on the performance of the ceramics. Chemical interactions between the ceramics and the beryllium also appear to be negligible. The tritium inventories determined for the SIBELIUS lithium zirconate and lithium oxide samples were extremely low, at 7.8 M B q / g (0.02 wppm) and 35.5 M B q / g (0.096 wppm). While the inventories for the orthosilicate and aluminate are about two or- ders of magnitude higher, 910 and 1844 M B q / g (2.6 and 5.2 wppm), they are still low and below levels which would cause safety problems.

264 J.P. Kopasz et aL /Journal of Nuclear Materials 219 (1995) 259-264

References

[1] N. Roux, M. Briec, M. Bruet, T. Flament, C. Johnson, M. Masson, A. Terlain and F. Tournebize, J. Nucl. Mater. 179 (1991) 827.

[2] N. Roux, J.J. Abassin, M. Briec, D. Cruz, T. Flament and I. Schuster, J. Nucl. Mater. 191 (1992) 168.

[3] D.L. Baldwin, O.D. Slagle and D.S. Gelles, J. Nucl. Mater. 179-181 (1991) 329.

[4] D.L. Baldwin and M.C. Billone, Diffusion/desorption of tritium from irradiated beryllium, presented at the 6th Int. Conf. on Fusion Reactor Materials, Stresa, Italy, September 27-October 1, 1993.

[5] W. Dienst, D. Schild and H. Werle, Tritium Release of Li4SiO4, Li20 and Beryllium and Chemical Compatibil- ity of Beryllium with Li4SiO4, Li20 and Steel (SIBELIUS Irradiation), Kernforschungszentrum, Karlsruhe, Report KIK-5109, December 1992.

[6] O.D. Slagle, T. Takahashi, F.D. Hobbs, K. Noda, D.L. Baldwin, G.W. Hollenberg and R.A. Verrall, Postirradia- tion Examination of BEATRIX-II, Phase I, to be pub- lished in J. Nucl. Mater.

[7] R.A. Verrall, J.M. Miller, l.J. Hastings, D.S. MacDonald

and D.H. Rose, in Advances in Ceramics: Fabrication and Properties of Lithium Ceramics, vol. 27, eds. G.W. Hollenberg and I.J. Hastings (American Ceramic Society, Westerville, OH, 1990) p. 299.

[8] H. Kwast, A. Kout, R.P. Muis, M.P. Stijkel, A.J. Flipot, and J.D. Elen, Fusion Eng. Des. 8 (1989) 359.

[9] T. Kurasawa, H. Yoshida, H. Takeshita, H. Watanabe, T. Miyauchi, T. Matui, H. Umei and Y. Naruse, J. Nucl. Mater. 133&134 (1985) 196.

[10] R.G. Clemmer, P.A. Finn, R.F. Malecha, B. Misra, M.C. Billone, D.L Bowers, A.K. Fischer, L.R. Greenwood, R.F. Mattas, S.W. Tam, R.B. Poeppel, G.T. Reedy, I.T. Dudley, F.F. Dyer, E.D. Clemmer, J.S. Watson, P.W. Fisher, J.R. Conlin, R.L. Childs, J.L. Scott, R.M. Arons and A.E. Scandora, The TRIO Experiment, Argonne National Laboratory, Report ANL-84-55 (1994).

[11] H. Kwast, R. Conrad, R. May, S. Casadio, N. Roux, and H. Werle, The behavior of ceramic breeder materials with respect to tritium release and pellet/pebble me- chanical integrity, presented at the 6th Int. Conf. on Fusion Reactor Materials, Stresa, Italy, September 27- October 1, 1993.

[12] N. Roux, personnal communication, April 1994.

performance of ceramic breeder materials in the sibelius experiment

Documents