canadian ceramic breeder technology: recent results

E L S E V I E R Fusion Engineering and Design 27 (1995) 297-306

Fusion Engineer!rig and Design

Canadian ceramic breeder technology: recent results

P. Gierszewski a, H. Hamilton b, j. Miller b, j. Sullivan b, R. Verrall b, j. Earnshaw c, D. Ruth d, R. Macauley-Newcombe e, G. Williams f

~Canadian Fusion Fuels Technology Project, 2700 Lakeshore Road West, Mississauga, Canada bAECL Research, Chalk River Laboratories, Chalk River, Canada

CTrent University, Physics Department, Peterborough, Canada dUniversity of Manitoba, Department of Mechanical Engineering, Winnipeg, Canada

eMcMaster University, Department of Engineering Physics, Hamilton, Canada fSpectrum Engineering Corp., 544 MeDonnel Street, PO Box 687, Peterborough, Canada

Abstract

Pebble bed ceramic breeders have been under development in Canada for over ten years. The goal is to fabricate and characterize these materials for use in engineering test reactors and in subsequent fusion power reactors. The program emphasis is on 1.2 mm diameter LizZrO3 and Li2TiO3 pebbles. Practical use of these pebbles requires a mass-production fabrication process, and characterization of the pebble beds with respect to bed behaviour and irradiation effects. This paper summarizes the relevant work within Canada since 1991.

The fabrication process presently used is suitable for mass production, and is in the process of being transferred to industry. Thermal cycling tests have been conducted on zirconate and titanate pebbles under both laboratory and "engineering" conditions. Cycling reduces the pebble strength, although there are indications that different fabrication conditions produce more robust pebbles. This is an active area of work.

Single-size lithium zirconate pebbles have been well-characterized in terms of the bed thermal conductivity and purge gas pressure drop. Recent results include measurement of thermal conductivity from 100 to 1200 °C (and 0 - 2 bar), and of purge gas pressure drop as a function of porosity. Binary beds have also been studied, using steel or lithium zirconate smaller pebbles.

Extensive irradiation testing of the as-fabricated ceramic is a critical factor in their acceptance. Lithium zirconate has been characterized under several European irradiation tests, and 1.2 mm lithium 'zirconate pebbles have been tested to 5.2% lithium atom burnup and over 250-1000 °C in the BEATRIX-II and CRITIC-2 purged-capsule experiments. Tritium release is rapid even at low temperatures, with no effects of burnup seen. The pebble bed temperature has been consistent with model predictions, and stable under irradiation. Post-irradiation anneal tests of lithium titanate show good tritium release. Post-irradiation examination of the BEATRIX-II lithium zirconate pebbles is just beginning.

Reference blanket designs have been developed based on breeder-in-tube geometry. Engineering-oriented tests have been carried out on large-volume (41) and long-pin (3 m) geometries, to characterize the behaviour of the pebble beds under more realistic conditions.

The results of the work described here, and related tests within the world fusion community, continue to support the use of these ceramic breeder pebbles in fusion reactor blankets.

Elsevier Science S.A. SSDI 0920-3796(94)00300-9

298 P. Gierszewski et al. / Fusion Engineering and Design 27 (1995) 297-306

1. Introduction

Pebble bed ceramic breeders have been under development in Canada for over ten years. This paper summarizes the results since 1991 [1]. The goal is to fabricate and characterize these materials for use in engineering test reactors and in subsequent fusion power reactors. Practical use of these pebbles requires a mass-production fabrication process and characterization of the pebble beds with respect to bed behaviour and irradiation effects.

The emphasis of the Canadian program has been on Li2ZrO3, due to its low tritium retention, and good combination of thermal/mechanical stability. The major effort was on irradiation tests, including the BEATRIX- II, CRITIC-2 and EXOTIC-7 tests of Li2ZrO3 pebbles, on improving the fabrication parameters, and on char- acterizing behaviour of pebble beds.

Recently, our work has also included an alternative low-activation ceramic breeder material--LizTiO3. The one prior measurement had reported reasonable thermal conductivity and tritium release [2]. Furthermore, a comparative evaluation noted no substantial disadvan-

tages for the material, suggesting to us that it could be a good compromise among all the requirements for a ceramic breeder [31.

In order to ensure that the R&D was directed at a useful product, the use of 1.2 mm ceramic breeder pebbles was explored in design studies, culminating in a reference driver blanket concept for NET/ITER [4,5]. Although the reference designs are based on single-size pebble beds, there are several reasons to consider binary beds of ceramic breeder or of mixed ceramic multiplier, including improved breeding and reduced local peak burnup. In exploring the usefulness of binary beds, scoping tests were conducted in order to bench- mark models for use in design evaluations.

The results of these studies are summarized in this paper.

2. Blanket design

In our recent studies, we have focused on a fission pin or BIT geometry, surrounded by a cooling annulus [4-6]. The space between the tubes is filled with beryl-

Table 1 Recent pebble bed blanket designs

NET [4] ITER EDA [5] CFFTP pilot [6]

Fusion power (MW)

Blanket purpose

Coolant

Coolant conditions

Breeder

Breeder conditions

Neutron wall load (MW m -2)

Neutron fluence (MW yr m -2)

Power tolerance (%)

Tritium breeding ratio

Breeder volume (m 3)

Breeder temperature (°C)

Purge pressure drop (kPa)

Breeder tritium (gT)

1000

Partial tritium breeding

Water

230-260 °C or 60-90 °C, 6.3 MPa

Li2ZrO3

1.2 mm diameter 55% smear density 65% packing

0.8 avg.

1 avg.

+20

0.57 3-D @ 50%6Li

22

850 max.

75

20 ("hot" coolant)

1500 20

Partial tritium Reactor blanket breeding test

Water Helium

160-200 °C, 2.5 MPa 200-400 °C, 5 MPa

LizZrO3 Li2ZrO3

1.2 mm diameter 1.2 mm diameter 55% smear density 55% smear density 65% packing 65% packing

1.0 avg. 0.25/0.35 avg/peak

1 avg. 0.25/0.35 avg/peak

+ 30 CW +230 for 10 s

0.95 3-D @ 50%6Li 0.35 3-D @ 50% (no ports) 6Li (outboard

blanket only)

- - 0.14

840 max. 1000 max.

75

140 0.2

P. Gierszewski et al. / Fusion Engineering and Design 27 (1995) 297-306 299

lium pebbles. Even though the breeder geometry is comparatively simple, the pin curves in at least one plane and vary in pin diameter across the blanket, so are easier to fill with pebbles than with close-tolerance pellets.

The results from these design studies are summarized in Table 1. In the NET/ITER design, Li2ZrO3 pebbles in a breeder-in-tube (BIT) type of geometry were used with low-pressure water cooling to provide a conserva- tive driver blanket design. The same mechanical configuration, without any breeder present and/or with the beryllium pebbles replaced by steel pebbles, would be able to provide initial machine shielding, with these driver blankets substituted as basic confidence was developed in the thermomechanical design. The CFFTP pilot plant considered a design for a minimum cost tokamak to provide nuclear testing of core components. Reactor-relevant breeding blankets were placed on the outboard, with only shielding blankets on the inboard. In this case, the same BIT type of blanket was considered as in the NET study, but with helium cooling and ferritic steel structure.

3. Fabrication

The basic fabrication process for lithium zirconate powder and 1.2 mm pebbles has been previously described [1]. The only major change has been in the mixer stage prior to extrusion. As a result of this and other minor changes, pebble production is now satisfac-

torily reproducible and the resulting pebbles are more uniform and denser than those available in 1991. Scop- ing studies on agglomeration and sol-gel methods for making smaller zirconate pebbles were unsuccessful. Lithium metatitanate powder and 1.2 mm pebbles are fabricated by an essentially identical process. Qualita- tively, it is observed that the titanate is easier to handle than the zirconate and appears to be more stable during storage in the green pebble state.

At the back end of the blanket cycle, the used ceramic breeder must be reprocessed to recover the largely unburnt 6Li. One possible route is suggested by our tritium studies, in which ceramic breeder specimens are dissolved for analysis of residual tritium content. Typically ~ 300 mg specimens are placed in aqua regia (3:1 nitric:hydrochloric acid) at ,~ 10-20 g 1-1, and the solution is refluxed for about 4 h. This results in 98% of the lithium dissolving into the solution. For lithium metatitanate, specimens are refluxed for 4 h (at ~ 1 g 1-1) in solutions of 95% aqua regia/5% HF, resulting in about 90% of the lithium going into solution.

Previously, powder and pebble fabrication was largely carried out at AECL Chalk River Laboratories. Recently, various aspects of the fabrication process have been tested with industry in order to develop a large-scale production capability. For example, in one series of tests, selected companies prepared batches of lithium metazirconate and metatitanate powder. These powders were examined and pressed into pellets or pebbles at AECL. The results are summarized in

Table 2 Comparison of powders fabricated by various industrial sources

Powder Supplier Particle Particle size, size, SEM Sedigraph (~tm) (~tm at 50

mass%)

Surface Sinter Pebble Pebble area pellet density crush (m 2 g-1) density a (%TD) load (N)

(%TD)

Lithium metazirconate

Lithium metatitanate

AECL - - - A <15 3.7 B < 20 2.8 C <15 7.2

AECL - - - - A <15 6.0 B <20 3.1 C-1 <5 9.0 C-2 < 15 7.8

4.3 - - 82 i0 1.5 88.6 - - - - 2.2 86.2 - - - - 1.5 86.8 - - - -

2.9 - - 82 36 0.5 86.2 77 17 5.2 87.5 87 40 2.0 85 c __ __ 0.8 86.4 c __ __

a Mainly monoclinic phase under XRD. Pellets/pebbles b Mainly monoclinic phase under XRD. Pellets/pebbles c Based on one pellet only.

sintered at 1225 °C/4 h. sintered at 1300 °C/4 h.

300 P. Gierszewski et al. / Fusion Engineering and Design 27 (1995) 297 306

Table 2. Lithium zirconate powders were produced with generally similar characteristics. Lithium titanate powders showed considerable variation in surface area, which appeared to influence the density and strength of the sintered pebbles.

Mechanical strength under thermal cycling is presently the major fabrication issue. Tests have been carried out with 1.2 mm pebbles at AECL, and at Trent University and Spectrum Engineering. For example, Li2ZrO 3 and Li2TiO 3 pebbles fabricated under different extrusion arid sintering conditions were cycled between 200 and 1000 °C in air at a ramp rate of 150 °C h - 1 for up to 60 cycles. The pebbles sintered at the lowest temperature generally had the highest crush strengths, smallest grain sizes and lowest densities. Typical results for Li2ZrO 3 pebbles were 12-15 N Weibull crush strength at 76-78% density and ~ 7 gin grain size, and 4 -5 N at 82-84% density and 40 gm grains. The Li2TiO 3 pebbles were stronger, typically 35-40 N at 79-85% density. For the stronger pebbles, both the density and strength initially increased up to about 15 thermal cycles, before subsequently decreasing. Possible explanations for this weakness are: anisotropic thermal expansion of the crystals, grain growth above a critical

grain size, or as a result of interaction with air (fabrication and testing have been carried out in air) [7]. The best approach to improving strength will depend on which of these dominates.

4. Bed properties

4.1. Bed packing density

A key characteristic of pebble beds is the overall effective smear density. This is a combination of the pebble internal density, and their packing density into the bed. The internal density is a result of the fabrication process. For packing, the general rule of thumb is that the minimum container dimension should exceed 10 times the pebble diameter in order to ensure a reliable packing fraction of around 63%. However, Li2ZrO 3 pebbles fabricated by the AECL process have shown consistently good packing behaviour, including packing fractions ranging up to 70% (Table 3). This good packing is reproducible, has been obtained by different groups, and is readily obtained by simple pouring and moderate tapping, even in long ( ,~ 3 m) pins.

Table 3 Packing fraction experience with 1.2 mm Li2ZrO 3 and Li2TiO 3 pebbles

Bed cross-section Bed length Pebbles Smear density Pebble density Packing fraction (mm ID/OD) (ram) (%) (%) (%)

Trent U. 1992 30/84 500 1.2 mm 52 82 63 Li2ZrO 2

BEATRIX-II 2.3/13.2 10 4 1.2 mm 52 80 65 1991 Li2ZrO2

CRITIC-2 1993 10.4/38 90 1.2 mm 53 82 65 Li2ZrO3

EXOTIC-7 0/7.9 84 1.2 mm 49 84 58 simulation 1993 0.94/7.9 Li2ZrO 3 48 57

U. Manitoba 3.8 ~ 600 1.2 mm 43 84 51 1991 6.7 Li2ZrO 3 52 62

9.6 57 68 11.6 56 67 12.6 59 70 25.2 57 68

UCLA 1993 25.4 50 1.2 mm 51 82 65 Li2ZrO3

Spectrum Eng. 12.7 ~ 3000 1.2 mm 58 82 1993 LizZrO3

AECL 1994 26 100 1.2 mm 51 79 Li2TiO3

70

65


Part of the good packing appears to be due to geometry. For example, a thin wire down the centre of a cylinder (e.g. thermocouple) need not reduce the packing to that expected for a tube of half diameter. In this case, there is still considerable freedom for the pebbles to arrange themselves around the central tube, and the overall affect on packing is small.

Secondly, part of the good packing seems to be due to the nature of the pebbles themselves. The combination of pebble shape, material and size distribution associated with the AECL fabrication process seems to allow the pebbles to flow readily together into packings that have significantly higher densities than that expected for randomly packed perfect spheres.

Thirdly, it should be noted that the zirconate pebble densities are not precisely known and there is some variability between batches, as we have been exploring fabrication parameters. This leads to some error in calculating the packing fraction from measurements of smear density.

Overall, we conclude that when sufficient space is allowed for the 1.2 mm Li2ZrO 3 pebbles (i.e. diameter ratios greater than 9), smear densities of 53-59% are reproducibly obtained with modest amount of vibration. Similar results are expected for Li2TiO3. For packing in long pins, the pebbles can be poured into the annular region around a rod temporarily inserted into the pin and slowly withdrawn as the pin is filled. This minimizes the drop impact force on the pebbles.

Tests have also been conducted with binary beds, where the small pebbles were either ~ 0.2 mm Li2ZrO 3

Table 4 Packing fraction experience with binary pebble beds using 1.2 mm

pebbles supplied by JAERI or 0.1-0.2 mm steel pebbles. These binary bed test results are given in Table 4 for cases where uniform packing was achieved. We conclude that the smaller pebbles should be less than about 0.17 mm diameter in order to pack easily within the larger 1.2 mm pebbles. Such small pebbles would readily flow down through the interspace between the larger pebbles. Otherwise, the small pebbles tended to bridge and jam among the large pebbles and achieving uniform packing required intensive layering effort lead- ing to low overall packings.

4.2. Bed thermal conductivity

Measurements were completed on the effective thermal conductivity of 1.2 mm Li2ZrO 3 pebble beds, over the full range of temperature and gas pressures of interest. These tests were carried out in three separate test runs, with intermediate modifications to the test facility. Fig. 1 summarizes the results. The newer results are consistent with the previously reported values [1]. The gas pressure dependence is consistent with that reported for A120 3 [1], and by other groups.

Measurements on these pebbles were also made by UCLA in air and helium, at low temperature [8]. In the most accurate test of thermal conductivity, the results were consistent with extrapolations of Trent Univer- sity's data to the lower temperature. In a second but less accurate test apparatus, a higher thermal conductivity was reported. Of more interest in the latter test was the measurement of the effect of external load on

Li2ZrO 3 pebbles

Column Length Large Small Smear Large Small diameter (cm) pebble pebble density pebble pebble (ram) diameter diameter (%) packing packing

(mm) (gm) fraction a fraction a (%) (%)

U. Manitoba 1992

U. Manitoba 1993

10.3 60 1.2 <250 NA 59 23 Li2ZrO 3 177-250 b b

149 177 63 20 106-149 64 21 74--106 63 21 <74 64 22 steel

9.52 80 1.2 150-250 54 48 17 Li2ZrO3 Li2ZrO 3

Large Li2ZrO 3 pebbles estimated to be 84% dense; small Li2ZrO 3 pebbles estimated to be 86% be almost 100% dense. u Small pebbles did not mix to form a suitable binary bed.

dense; steel pebbles estimated to


2.5

2

1.5

1

0 . 5

0

o

o ~ . . . _ ~ j >~ • •m

. ~ . ~ . ~ . . . . . . . . . . . . . .

I I I I I

200 400 600 800 1000 1200

Temp, *(2

• Trent 1

Trent 2

* Trent 3

o UCLA

. . . . . . . . 82%TD Li2ZrO3

. . . . . Helium gas

- - Schlunder model

Fig. 1. Thermal conductivity of 1.2 mm Li2ZrO 3 pebble beds in 1 bar helium. Conductivity of helium gas and of breeder material alone are also shown for comparison.

the pebble bed thermal conductivity. Specifically, no effect was found within experimental error ( ~ 10%) for loads up to 1.3 MPa for the bulk bed or near-waU conductivity in helium gas. This is consistent with theo- retical expectations for these hard, low-conductivity materials.

The ceramic breeder pebble data has been compared with a variety of models, including Hall and Martin, Schlunder et al. and others [9]. Although several models are capable of describing the single-bed data, a modified Schlunder model was found to give the best overall match to data for a range of sizes, conditions and materials of fusion interest. The good correlation of the Li2ZrO3 data with the modified Schlunder model is also shown in Fig. 1.

For 1.2 mm Li2ZrO3 pebbles ( ~ 82%dense, 63% packing) in 0.1 MPa helium, the bed thermal conductivity data can be fitted by

kbed(W m - 1 K -1) = 0 . 6 6 + 1.17 x 1 0 - 7 Z ( ° C ) 2"2 (1)

For 1.2 mm Li2TiO3 pebbles under similar conditions, the bed conductivity has not been measured, but can be estimated from the bulk conductivity [5] and Schlun- der's model [10] as

kbed(W m ~1K -~) = 0 . 6 2 + 5 . 5 × 10-4T(°C) (2~

Near the container walls, the thermal behaviour of the bed will vary from the above effective conductivity model. This can be treated as either an interface heat transfer coefficient, similar to those used in considering pellet-clad gaps, or as an effective bed near-wall con-

ductivity over the affected region some distance from the wall. We think the latter is more realistic. Conse- quently, we define a near-wall conductivity knw for the region that is most affected by the altered packing near the wal l - -wi thin one pebble radius. As it is a comparatively small effect in most experiments, there is considerable scatter in the data. The Trent University thermal conductivity data imply knw values of about 0.2-0.4 W m-~ K-1 for 1.2 mm Li2ZrO3 pebbles in helium gas.

4.3. Bed gas permeability

Pebble bed breeders provide a stable pore network for uniform, distributed flow of the purge gas that is not sensitive to cracking or shifting, as may occur with pellets. However, the more tortuous path in pebble beds may result in higher purge pressure drops. This path length increases as the pebble size gets smaller, and the resulting increased pressure drop contributes to the lower practical limit on pebble diameter of around 0.1 mm [1].

Measurements have been made for gas permeability through single and binary beds containing 1.2 mm Li2ZrO 3. The pressure drop was measured at room temperature, with helium or air flowing at low speeds. The resulting pressure drop Ap can be described by separating the bed contribution from the external conditions as follows:

Ap = #vL/kp (3)

where ~t is the gas viscosity, v is the superficial gas velocity, L is the bed length and kp is the intrinsic bed


permeability. This permeability is a characteristic of the packed bed, and may be obtained directly from experiments or estimated by a model such as the Carman- Kozeny model:

k p = E 3 / 5 S 2 ( 1 - - E) 2 (4)

where E is the bed porosity and S is the bed surface area per unit volume; for uniform spheres, S = 6 ( 1 - E)/D, where D is the sphere diameter.

In the experiments, the 1.2 mm LizZrO 3 single-size beds were packed over tube diameters from 3.8 to 25 ram, and a small range of packing fractions (loosely poured to vibration compacted) at each diameter. The measured bed permeability is somewhat higher than the Carman- Kozeny equation, and can be fitted by the following expression for tube diameters larger than 9 ram:

kp = 13622e 3/( 1 - E) 2/, m 2 (5)

Binary beds of ceramic breeder and steel (simulating beryllium) were tested at a fixed tube diameter (10.3 ram), but for a range of steel pebbles ranging from

0.04 to ~ 0.2 mm diameter. Based on these experimental results, for well-mixed binary beds of 1.2 mm Li2ZrO 3 pebbles and steel pebbles, the bed permeability can be estimated by

kp = (1630 x D22) x ~ 3 / ( e I - - e)zl~m z (6)

where D2 is the smaller pebble diameter in mm, E is the binary bed porosity, and E1 is the porosity of the bed with the larger pebbles only.

0.15-0.25 mm Li2ZrO3 pebbles were supplied to CFFTP under a collaborative effort with JAERI. The pressure drops of single and binary beds using these pebbles were also measured. For a 0.2 mm pebble bed alone, the permeability was kp ~ 22 i.tm 2. This compares with 29 gm 2 obtained by scaling the 1.2 mm results by the diameter ratio (and assuming similar packing). A good binary bed could not be obtained because of the pebble diameter ratio. For a reasonably well-mixed binary zirconate bed, but with a low 65% overall packing, the measured bed permeability was 14 ~tm 2, compared with 100 ~tm 2 estimated from the steel-zirconate binary bed equation. This low effective permeability may be due to the non-uniformity of packing in this bed.

5. Nuclear testing

5. I. Lithium metazirconate

A considerable amount of irradiation data is now accumulating for Li2ZrO 3. Irradiation tests by other

groups include FUBR, EXOTIC-4, 5, 6, TRIDEX-5, MOZART, SIBELIUS and COMPLIMENT. In addi- tion, CREATE post-irradiation anneal tests have been conducted [11]. In general, it is observed that lithium zirconate releases tritium readily even at relatively low temperatures (e.g. 300 °C). The experimental data has been interpreted as either diffusion-controlled or second-order desorption, but these are not firm conclu- sions. Measured effective activation energies range from 66-120 kJ mo1-1.

AECL-fabricated lithium zirconate pebbles have been tested in the BEATRIX-II [12] and CRITIC-2 [13] irradiations, and in the EXOTIC-7 test in 1994. Al- though the irradiated pebbles have not been examined yet, some general observations can be made.

BEATRIX-II Phase II was in FFTF, the high-flux fast-neutron test reactor at Westinghouse Hanford (USA). High-enriched (85% 6El) 1.2 mm pebbles were tested in various H e - H 2 purge gases (0-0.1% Ha) to a burnup of 5.2% lithium. The bed had a temperature difference of 400 °C surface to 1100 °C centre, in good agreement with predictions based on the measured bed thermal conductivity. The temperature remained constant throughout the irradiation, after allowing for small decreases in heat generation as the neutron flux decreased. This is interpreted as indicating that the pebble bed showed reasonably good stability. The tritium release showed a small apparent decrease with decreasing H 2 content, but was otherwise constant. Most of the tritium appeared to be released as H T - - l i t - tle moisture content was observed in the purge after an initial pulse during startup.

CRITIC-2 was a similar test in NRU, a thermal reactor at AECL Chalk River Laboratories, with a larger amount of pebbles, lower temperatures and smaller burnup. The temperature difference between surface and centre was about 600 °C, and the surface temperature could be varied from as low as 200 °C nominal (there were temperature differences circumfer- entially and longitudinally). The on-line release data of this experiment has not been analysed as the experiment was just concluding at the time of writing. First indications are that tritium release was nearly as high at the lowest temperature of operation (200 °C surface) as at the highest temperature (350 °C surface). The breeder was exposed to ~ 90 temperature cycles from operating temperature to coolant temperature as a result of reactor operations. The cumulative irradiation time was about 300 days.

The pebble burnup database extends to 5% bur- n u p - - I T E R end-of-life conditions. The EXOTIC-7 test, which recently started, should provide data up to


10% burnup. There is also some data, however, that may approximate the effects of full burnup. In this "accidental" experiment, Li2ZrO3 pebbles appear to have shorted heater wires due to the increasing electrical conductivity of lithium zirconate at high temperatures ( > 1100 °C). The pebbles were examined afterwards and found to be composed of ~ 80% zirco- nia (i.e. the Li20 had been evaporated off). Interest- ingly, the resulting pebbles were more porous but had a stronger crush strength than the original lithium zirconate pebbles, probably reflecting the known high strength of ZrO2 ceramic.

5.2. Lithium metatitanate

Tritium release was examined in post-irradiation an- nealing tests [2,14,15]. Samples of Li2TiO 3 spheres, sintered at 1498 K and 1673 K were vacuum annealed in a quartz ampoule and irradiated. Fig. 2 shows tritium desorption data for the Li f f iO 3 spheres heated at a linear rate. Even with pure He sweep gas, a significant fraction of the tritium inventory is released below 650 K. The results with ramped temperature tests compare

favourably with those for similar lithium zirconate pebbles. The shape of the release curve is believed to indicate tritium release from several states or surface sites, rather than diffusion-limited release. An activation energy of 35 kJ mo1-1 was inferred [14] assuming first-order desorption controlled the maximum tritium peak, with He/0.1% H2 sweep gas. Tritium was found to be released largely as HTO. This is consistent with CREATE tests on lithium zirconate, but in contrast with CREATE tests on LiA102 and Li20.

Lithium titanate sintered material has also been examined using ion beam analysis after D 2 loading. The results yield an upper bound for D 2 solubility, since surface adsorption cannot be separated, of ~3 x l O - T e x p ( + 3 O k J m o l - 1 / R T ) mol fr./Pa °5 over 300- 500 °C and 2-100 kPa.

6. Engineering tests

In general, the experiments described earlier were intended to measure specific properties, such as bed thermal conductivity or tritium release. However, even

1 4 -

1 2 -

1 0 -

8 -

f i $

i / ! I

i

i

I Li2ZrO a

,. Li2TiO 3 (1498 K)

,"-"-"-""" Li2TiO 3 (1673 K)

, / \ . , -..

e¢ , ¢

_ - - i | " ' '

0 I I I I I I

300 400 500 600 700 800 900

T e m p e r a t u r e (K)

Fig. 2. Tritium release from Li2TiO 3 and LizZrO 3 pebbles during out-reactor thermal ramp tests (linear heating rate of 2 K rain-l, pure He sweep gas).

P. Gierszewski et al. / Fusion Engineering and Design 27 (1995) 297 306

Table 5 Summary of "engineering" tests of 1.2 mm Li2ZrO3 pebble beds

305

ITER CRITIC-2 ENEA S p e c t r u m BEATRIX-II Trent

Pin ID (ram) 16-29 11/38 6/20 12 2/13 25/97 Pin length (m) 10 0.1 2 3 0.1 0.6 Pin volume (L) 2-6 0.09 1 0.4 0.014 4.5 Density (g cm -3) 2.3 2.15 - - 2.4 2.14 - - Temperature range ( ° C ) 200-800 200-800 200 800 50-680 450-1100 100-1200 Max heating (W cm -3) 22 6 25 a 9 a 50 - - Max TPR (Ci cm -3 d l) 0.017 0.015 0 0 0.2 0 Purge gas He + 0.1% H 2 He+ 0.1% H 2 He Air He + 0.1% H 2 He HT pressure (Pa) 15 1 0 0 5 0 Clad (ram) 19 316L 3 Inconel 1.0 316L 1.6 304 3 8 Inc 600 Coolant Water Water Water Air Sodium Air Cycles (No.) up to 105 ~90 - - 65 9 11 Burnup (at.%Li) 4.1 ~0.5 0 0 5.2 0 Operating time (d) 365 ~ 300 - - 3.5 200 ~ 100 Status Design Complete Built Complete Complete Complete

a Approximate equivalent volumetric heating rate.

after the base properties have been characterized, it is still necessary to conduct "integrated" testing that in- corporates blanket relevant geometries and operating conditions. Table 5 summarizes the status of such tests on 1.2 mm Li2ZrO3 pebbles, in comparison with conditions expected in an 1TER driver blanket using these pebbles [5]. More details on these experiments are given below.

Trent University: although primarily thermal conductivity tests, these tests studied a large volume pebble bed for an extended period. In this large diameter assembly, the pebbles were found to settle quickly into a stable compact state, with about 5% less volume than the as-poured packing. Some dust formation was noted, but it was suspected to be primarily from the pebble removal process. No pebble-to-pebble sintering was observed, except if the system was not pre-baked or if the central pebbles were overheated.

BEATRIX-II: in the Phase II irradiation, one capsule contained Li2ZrO 3 pebbles. Significantly, these pebbles operated up to 5% burnup with a stable temperature profile. The heating and tritium production rates were higher than those expected in ITER. The irradiation has been completed, but post-irradiation examination has not started.

Spectrum Engineering: the purpose of this test was to explore pebble bed behaviour in a long-pin (BIT) geometry. Packing the bed was not difficult. The bed was thermally cycled in air, using external heaters and purge gas cooling. Although differential axial thermal expan-

sion of 14 mm was expected, none was observed. On disassembly, regions in the pebble bed had formed bridges that did not release until prodded with a wire, but there was visually no sense that the pebbles had formed necks. About 15% of the pebble bed mass had degraded into smaller pieces, with about 10% smaller than about 35 gm (i.e. few grains). There was no indication that these particles had settled down to the bed bottom. These pebble fracturing observations are consistent with those observed in unconstrained beds- - that is, no effect of the long-pin geometry was seen.

ENEA: another long-pin test is planned in collabora- tion with ENEA. Specifically, a 2 m long water-cooled pin test assembly has been built. Heating is to be provided by an internal heater wire. This assembly is more reactor-relevant in that the temperature profiles are more representative and a thin clad will be used. Therefore the interaction between the clad and the pebbles should be a significant aspect of this test (e.g. pebble settling, clad distortion).

CRITIC-2: in the largest pebble bed in-reactor test to date, 200 g of LizZrO 3 are being irradiated. The particularly significant features of this test are that the temperatures across the pebble bed correspond very closely to those for the ITER driver blanket design, with edge temperature as low as 200 °C. Therefore the final analysis of tritium inventory should be directly relevant to ITER conditions. The tritium production and heating rates, in conjunction with the BEATRIX-II results, bracket the ITER range.


7. Summary and plans

Since our last summary paper, a considerable amount of data has been obtained on 1.2 mm Li2ZrO3 pebbles. The results are consistent with those reported by other groups on pebbles of zirconate and of other materials [10], and provide a good basis for blanket design. The extent of nuclear and blanket module testing is also considerable, partly because the reference BIT configuration is readily tested in both in- and out-reactor facilities.

The major issue is pebble strength under thermal cycling, and quantifying the tr i t ium inventory. The immediate plan then is to support the irradiation, data analysis, and post-test examination of lithium zirconate pebbles from the BEATRIX-II , CRITIC-2 and EX- OTIC-7 experiments. On the fabrication side, the major task is to improve the pebble strength under thermal cycling.

Lithium metati tanate continues to look like an inter- esting alternative material. We will continue to improve its database and, in the longer term, prepare for an i r radia t ion test of lithium titanate pebbles.

Acknowledgements

These results were obtained with the efforts of many people. We particularly wish to thank M. Tillack and F. Tehranian at U C L A for their measurements of thermal conductivity, H. Yoshida and M. Enoeda at JAERI for supplying 0.2 mm lithium zirconate pebbles for binary bed tests, and the BEATRIX-I I and CRITIC-I I teams for preparing and carrying out these major irradiation tests.

References

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[2] P. Finn, T. Kurasawa, S. Nasu, K. Noda, T. Takahashi et al., Solid oxide compounds--properties necessary for

fusion applications, Proc. IEEE 9th Symp. Engineering Problems of Fusion Research, Chicago, 1981, 1200.

[3] M. Abdou, A. Hadid, A. Raffray, M. Tillack, T. Iizuka et al., Modelling, analysis and experiments for fusion nuclear technology, Fusion Eng. Des. 6 (1988) 3.

[4] P. Lorenzetto, P. Gierszewski, S. Chiocchio, W. Daenner, G. Federici et al., A driver blanket for NET, Fusion Technology 1992, Proc. 17th Symp. Fusion Technology, Rome, 1992, p. 1414.

[5] P. Lorenzetto, P. Gierszewski and G. Simbolotti, A Eu- ropean proposal for a ITER water cooled solid breeder blanket, Fusion Eng. Des. 27 (1995).

[6] P. Gierszewski, G. Williams, J. Blevins, H. Brunnader, P. Cumyn et al., CFFTP fusion pilot plant scoping study, Proc. 1 lth ANS Topical Meeting Technology of Fusion Energy, New Orleans, 1994.

[7] H. Hamilton and J. Sullivan, The thermal cycling behav- ior of lithium titanate, Proc. 4th Syrup. Fabrication and Properties of Fusion Ceramics (96th Annual Meeting of the American Ceramics Society, Indianapolis, 1993).

[8] M. Tillack, University of California at Los Angeles, personal communication, 1994.

[9] W. Fundamenski and P. Gierszewski, Heat transfer cor- relations for packed beds, Proe. 10th Topical Meeting on Technology of Fusion Energy, Boston, 1992.

[10] P. Gierszewski, M. Dalle Donne, H. Kawamura and M. Tillack, Ceramic pebble bed developments for fusion blankets, Fusion Eng. Des. 27 (1995).

[11] J. Miller, S. Bokwa, D. MacDonald and R. Verrall, Tritium recovery from lithium zirconate spheres (Proc. 9th ANS Topical Meeting on Technology of Fusion Energy, Oak Brook, USA, 1990), Fusion Technol. 19 (1991) 996.

[12] R. Verrall, O. Slagle, G. Hollenberg, T. Kurasawa and J. Sullivan, Irradiation of a lithium zirconate pebble bed in BEATRIX-II Phase lI, Proc. 6th Int. Conf. Fusion Reac- tor Materials, Stresa, Italy, 1993, J. Nucl. Mater. 212- 215 (1994) 902.

[13] J. Miller and R. Verrall, Performance of a LizZrO3 sphere-pac assembly in the CRITIC-II irradiation experiment, Proc. 6th Int. Conf. Fusion Reactor Materials, Stresa, Italy, 1993.

[14] J. Miller, H. Hamilton and J. Sullivan, Testing of lithium titanate as an alternative blanket material, Proc. 6th Int. Conf. Fusion Reactor Materials, Stresa, Italy, 1993.

[ 15] J. Kopasz, J. Miller and C. Johnson, Tritium release from lithium titanate, a low activation tritium breeding material, Proc. 6th Int. Conf. Fusion Reactor Materials, Stresa, Italy, 1993.

canadian ceramic breeder technology: recent results

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