summary of experimental results for ceramic breeder materials

E L S E V I E R Fusion Engineering and Design 27 (1995) 154-166

Fusion Engineering and Design

Summary of experimental results for ceramic breeder materials

N. Roux a, G. Hollenberg b, C. Johnson c, K. N o d a d, R. Verrall e a CEA/CE Saclay, Saclay, France

c Battelle, PNL, Richland, WA, USA c Argonne National Laboratory, Argonne, IL 60439, USA

a Japan Atomic Energy Research Institute, Tokai-rnura, Naka-gun, Ibaraki-ken 319-11, Japan e AECL, Chalk-River, Ont., Canada

Abstract

Lithium-containing ceramics were quickly recognized as promising tritium breeding materials for fusion reactor blankets, particularly because of their safety advantages. Relevant material properties were investigated to evaluate further their suitability. An extensive R&D program complemented a conceptual blanket design activity either for near-term machines or for power reactors. All aspects were addressed: fabricability; overall properties (baseline, thermal, mechanical); compatibility with structures and beryllium; tritium release characteristics; irradiation behavior; activation; reprocessing; waste disposal issues. As a result of this investigation, lithium-containing ceramics are considered to be excellent tritium breeding materials.

1. Introduction

In the development of tritium breeding blankets for fusion reactors, lithium-containing ceramics were quickly recognized as promising tritium breeding materials. Their excellent thermal stability and chemical inertness indicate favorable safety characteristics which is a principal attribute of fusion power. Furthermore, no magnetohydrodynamic (MHD) effect is to be feared using these compounds.

An extensive international R&D program was established to confirm the suitability of these attractive materials. Issues closely related to the primary functions of the blanket, i.e. tritium breeding, tritium release and energy conversion, were addressed. Property characteristics necessary for design analysis, and for assessment of blanket concept performance, lifetime, reliability and

Elsevier Science S.A. SSDI 0920-3796(94)00328-9

safety were investigated. Thus, the research programs focused on fabrication, property measurements, compatibility with other blanket materials, tritium release behavior, irradiation behavior, activation characteristics, reprocessing, and waste disposal issues. As a result of this international collaboration, lithium oxide and the ternary ceramics (lithium-aluminate, zirconate, silicate and titanate) were identified as promising candi- dates.

The ceramic breeder research complemented conceptual blanket design activities. Two configurations of ceramics were considered, i.e. sintered bodies and pebble beds. Following the lead of the nuclear fission industry, blanket concepts using ceramics in pellet form were naturally envisaged. Thus, the EU DEMO BIT blanket concept features rows of breeder modules containing annular LiA102 or Li2ZrO3 pellets [1]. The

N. Roux et al. / Fusion Engineering and Design 27 (1995) 154 166 155

pebble-bed concept option is attractive, because it par- tially alleviates cracking concerns. Also, loading of pebbles into complex geometries can be easier than with pellets, and intimate mixing of ceramic breeder and neutron multiplier materials can be facilitated by mixing pebbles of the two materials. For example, in the latest version of the EU DEMO BOT design, pebbles of beryllium and Li4SiO 4 are used. The larger (2mm) beryllium pebbles fill about 60% of the bed space. Spaces between the pebbles are filled with a mixture of pebbles of Li4SiO 4 and beryllium with sizes of 0.1- 0.2 mm [2]. Likewise, the Japanese concept for ITER considers a layered pebble bed consisting of Li20 pebbles 1 mm in size separated from beryllium pebbles [3].

Blanket designs considered a lithium ceramic as a first option, based upon its attractive properties. Com- plementary materials research developed the necessary database and optimized the property characteristics as appropriate. Indeed, several properties of ceramics can be adjusted through material tailoring.

2. Experimental results and performance analysis

The experimental results recalled in this section illus- trate the breadth of knowledge acquired and highlight the excellent performance of ceramic breeder materials.

2.1. Fabrication

The fabrication of ceramic breeder materials repre- sents a determining step, because it governs their performance. Fabrication includes the preparation of powders and shaping to the required form, i.e. pellets or plates, and pebble beds. Pellets are porous-- typi- cally 75%-80% of theoretic density--so easing tritium release, whereas pebbles are dense and the porosity is that of the bed. The bed porosity can be adjusted through varying the number of pebble sizes and size ratios.

The importance of the microstructure (grain size, density and pore morphology) and impurities for properties and performance has long been recognized. Thus, there is a general trend for fine materials, because early in situ tritium release experiments indicated their better tritium release performance. In addition, they are expected to exhibit better mechanical behavior. Impurities should be kept at the lowest level, because they can have an adverse effect on properties such as compatibility behavior and activation characteristics.

Several preparation methods using liquid or solid routes were worked out and yielded powders with

satisfactory purity levels [4-17]. Batch-scale production (10-100 kg) was easily accomplished at KIN, Ceramics Kingston, Uranium Pechiney and Temav. Likewise, fabrication into the required configurations was successfully accomplished using processes well-proven in the ceramic industry, in general, and in the nuclear fission industry, in particular. The scalability of the fabrication processes was explored. For example, LiA102 pellets prototypical of the first row breeder modules of the EU DEMOFBIT concept were produced by cold uniaxial pressing and sintering by Temav- ENEA [18]. The pellets fulfil the microstructural specifications as defined in the material research, i.e. grain size of about 0.5 gm and 80% of theoretic density, and the geometrical specifications optimized in the design work, i.e. annular pellets with an aspect ratio (outside/inside diameter) of 1.8. Several hundred pellets were fabricated for laboratory tests and irradiation experiments. Likewise, Li2ZrO3 BIT pellets were produced in batches of hundreds by Uranium Pechiney [19,20], by uniaxial pressing of a spray dried Li2ZrO3 powder and sintering. Adjusting the process parameters allowed spanning of the 75%-85% theoretic density range while maintaining the grain size at about 1 gm to preserve the tritium release behavior. The process was transposed to the fabrication of LiA1Q pellets.

The fabrication of Li4SiO 4, Li2ZrO3 and Li20 pebbles was successfully accomplished and scaled up to batch level [15,21-23]. Thus, batch quantities of 10 kg were produced of Li2ZrO3 spheres 1.2 mm in diameter at AECL (by extrusion, tumbling and sintering), Li4SiO 4 pebbles 0.4-0.6 mm and 0.1-0.2 mm in size at Schott Glaswerke Mainz (using a melt spraying process), Li20 spheres l mm in size at Kawasaki Heavy Industries (using a rotating granulation method) and LizTiO 3 spheres at AECL. To date, fabrication scale- up has not raised any major problem; however, when industrial quantities become necessary, further development will be required and sufficient lead times will have to be provided. Attention must be paid to storage of the finished product because ceramic breeders--especially L i20 - -a re sensitive to H20 and CO2 contamination. Irreversible alteration of the materials may result in degradation of the material performance.

In parallel, improvement of the materials properties was pursued, through composition tailoring and microstructural tailoring. For example, improvement of Li4SiO 4 spheres' mechanical stability was obtained by the addition of SiO 2 [24]; lowering of the tritium release temperature of LiA102 was achieved by silicon substitu- tion on the LiA102 lattice [25]; and the thermal cycling

156 N. Roux et al. / Fusion Engineering and Design 27 (1995) 154-166

performance of Li2ZrO 3 pellets was enhanced by den- stiy adjustment [20].

2.2. Properties characteristics

A wide spectrum of property characteristics necessary for blanket design analysis were determined. An overview is given here.

The lithium atom density, which has an impact on the tritium breeding ratio (TBR) and lithium burn-up at end-of-life, should be as high as possible. In this respect, Li20 is the most attractive ceramic. However, this factor is not as critical as previously thought, since there is evidence that a beryllium neutron multiplier and 6Li enrichment are necessary to achieve a TBR greater than unity. DEMO blanket designs do achieve TBR > 1, even with ternary ceramics.

Stability of the lithium ceramics at high temperatures is obviously attractive from a safety perspective. In addition, the capacity for blanket operation at higher temperatures is beneficial to the thermal efficiency. The melting points for the candidate ceramics are all above 1250 °C, indicating good thermal stability. High temperature phases are considered so that physical changes cannot occur during operation. The vapor pressures of the compounds and their constituents are sufficiently low at the temperatures of interest so that no chemical change can occur. The total partial pressures over the ceramic remain below 10 .2 Pa at the anticipated blanket operating temperatures and even under limited up- side transient conditions, ensuring no significant material transport by the purge gas.

The thermal properties, i.e. thermal conductivity and linear thermal expansion, and their dependences on temperature and, for the thermal conductivity, on the microstructure were determined. The thermal conductivity of monolithic lithium ceramics is not very h i g h - - in the range 0.8-3.5 W m -1 K -~ at 600 °C for materials at 80% of theoretic dens i ty - -and this has to be ac- counted for in the design process. The pebble bed thermal conductivity and heat transfer coefficients between the walls and the pebble bed are important because they determine the operating temperatures and temperature gradients in the blanket.

The effective thermal conductivity of pebble beds was determined under several different combinations of pebble size and density, packing fraction, purge gas composition, pressure and flow rate. Measurements were made on LiaSiO 4 pebble beds with pebbles 0.4-0.6 mm in size [26], on mixed pebble beds with beryllium pebbles 2 mm in size and Li4SiO 4 pebbles 0.1-0.2 mm in size, and on beds with beryllium pebbles 2 mm in size and

mixed Li4SiO 4 and beryllium pebbles 0 .1 -0 .2mm in size [27].

The thermal conductivity and heat transfer coefficients for the mixed beds are considerably higher than those for LiaSiO 4 beds. The experimental results agree with the model predictions. Similarly, the thermal conductivity of 1.2 mm Li2ZrO3 pebble beds was measured between 300 and 1200 °C [28]. The results follow closely the model predictions and are consistent with the bed temperatures observed in irradiation tests. Inde- pendent measurements at 100 °C at UCLA are in rea- sonable agrement. Finally, the thermal conductivity of 1 mm Li20 pebble beds, as considered for the Japanese ITER design, was measured [29].

Though ceramic breeders have no structural role, their thermomechanical behavior is important from a cracking resistance perspective. Breeder cracking is un- desirable, since it alters the heat transfer characteristics and may cause tritium purge channel plugging, as a result of the transport and relocation of breeder frag- ments by the purge gas. Relevant materials characteristics were determined, such as elastic constants (Young's modulus, Poisson's ratio) and fracture strength [6,30- 33], as well as their dependences on the density, grain size and temperature. The database for the baseline, thermal and mechanical properties for the elastic constants and thermal creep of ceramic breeders was re- viewed and assessed as part of the I T E R - C D A activity [34]. Properties correlations were developed, and are reported in Ref. [35].

The thermomechanical behavior of both pellets and pebble beds was checked under thermal cycling [20,27,28,36,37]. The thermomechanical behavior of pellets for the EU DEMO BIT concept was checked out-of-reactor, under conditions of heat generation, temperature, thermal gradient, cycling temperature range, and coolant flow rate, pressure and temperature typical of those encountered in the BIT blanket design. Thus, a stack of pellets 1 m long (about 125 pellets), representative of those in the first row breeder modules, was tested in the facility designed at ENEA. In this facility heat generation in the breeder is simu- lated by a wire resistor placed in the central channel formed by the annular pellet stack. Thermal cycles of 200 s of burn and 70 s of dwell are applied. Varying the power supplied to the resistor results in varying stresses in the pellets. The hydraulics of a helium purge flowing in the central channel is continuously moni- tored to identify any impact of pellet fracture. The fraction of broken pellets is determined after a number of cycles and the mode of pellet fracture is observed [381.

N. Roux et aL/ Fusion Engineering and Design 27 (1995) 154-166 157

Temav-ENEA LiA102 pellets at 80% of theoretic density were cycled in the temperature range 500- 700 °C, showing no fracture after 10 000 cycles at an equivalent tangential stress of 22 MPa, 3% fracture after 20 000 cycles, with no further fractures after 10 000 additional cycles at 32 MPa. Early tests of relatively low density Li2ZrO3 BIT pellets produced both at CEA and by Uranium-Pechiney showed poor behavior. However, Li2ZrO3 pellets with higher densities, i.e. 76.8% of theoretic density (laboratory scale) and 83.5% of the theoretic density (industrial), cycled in the range 330-550 °C, performed well. No fracture was observed for the pellets with 76.8% of theoretic density after I0 000 cycles at 30 MPa; 14% fracture was observed after 10 000 additional cycles at 40 MPa. No fracture was observed for the pellets with 83.5% of the theoretic density up to 30 000 cycles at 47 MPa, and 39% fracture was observed after 10 000 additional cycles at 57 MPa, indicating that the higher the density is, the better is the behavior [20].

These tests demonstrated the good behavior of LiA102 and LizZrO 3 pellets, and emphasized the importance of fabrication routes and related materials characteristics for the thermomeehanical performance. Fracture occurred neatly along a diameter, suggesting that, if pellets do fracture in a breeder rod, then there should be no detrimental consequence on the purge flow. This condition was indeed verified in the test itself, because no change in the hydraulics of the purge flow was observed.

The thermal cycling behavior of pebbles 1 mm in size was investigated for Li20 with 92% of theoretic density and grain size 42/am, LizZrO3 with 80% of theoretic density and grain size 40 lam, and Li4SiO4 with 93% of theoretic density and grain size 40/am, [36] under ITER representative conditions, i.e. 400-800 °C and heating- cooling rates of 20 °C s -1 up to 2000 cycles. While the Li20 pebbles performed extremely well, the Li2ZrO3 and Li4SiO4 pebbles fractured significantly. In contrast, tests of Li4SiO 4 pebbles 0.4-0.6 mm in size containing 2.2% SiO2, produced by Schott Glaswerke, indicated good performance [24]. Tests of Li4SiO 4 pebbles 0.4- 0.6 mm in size cycled for 50 cycles of beryllium pebbles 2 mm in size mixed with Li4SiO 4 pebbles 0.1-0.2 mm in size cycled for 1000 cycles, and beryllium pebbles 2 mm in size mixed with Li4SiO 4 and beryllium pebbles 0.1- 0.2 mm in size cycled for 500 cycles also showed good behavior [27,28]. The latter indicated a critical rate of taemperature change with time of 50 °C s 1, which is significantly higher than the peak value expected in the DEMO blanket (about 20°Cs-1). Furthermore, no fragment came out of the bed under purge gas velocities

considerably higher than that anticipated in the concept option. Li2ZrO3 pebbles sintered under various conditions and, hence, exhibiting different densities and fracture strengths were cycled from 20 to 1000 °C [37]. The results point out the effect of microstructural characteristics on the pebble performance.

2.3. Compatibility behavior

The compatibility of ceramic breeders with other blanket materials is an important consideration with respect to safety and operating limitations. The compatibility behavior is sensitive to impurities. Numerous studies have been made of the compatibility with structural materials--particularly with austenitic and ferritic steels--both in vacuum and flowing atmospheres, with and without water addition [39-43]. Tests with impurity-free ceramics indicated a reasonably low interaction at the temperatures of interest, i.e. below 700 °C. The penetration depth in steels was established for all ceramic breeders as a function of temperature [35]. The reaction rates increase in the presence of moisture. The effect of irradiation was explored in the COMPLI- MENT experiment (maximum lithium burn-up of 1.4%) [33]. No irradiation-induced processes were observed. A comparison of the results with those in annealing tests under 1 Pa of H20 pressure shows that the latter offer a conservative simulation condition [40].

The compatbility of ceramics and beryllium was investigated because some blanket designs place the beryllium and ceramics in intimate contact. In laboratory tests, the interaction of beryllium with the ternaries was found to be negligible up to 650 °C [41,44-46]; it is larger for Li20 [36,46]. This is contrary to thermodynamic expec- tations, because there is a large free energy driving force for oxidation of the beryllium. However, it is assumed that a thin layer of beryllium oxide forms, which protects the metal from further oxidation. The impact that neutron irradiation may have on the oxidation kinetics of beryllium was studied in the SIBELIUS experiment [47]. In this experiment, the interaction of beryllium and lithium ceramics was studied in the mixed-spectrum Siloe reactor core during 2000 hours at 550 °C (average materials interface temperature) and He + 0.1% H2 purge. No irradiation effect was detected under the conditions explored at 6Li burn-up of 20%-40% for natural 6Li isotopic content (Li burn-up of 1.5%-3%) [48,49].

2.4. Tritium release and recovery behavior

In an operating fusion reactor, the tritium breeding blanket will reach a condition in which the tritium


release rate equals the production rate. The tritium release rate must be fast enough that the tritium inventory in the blanket does not become excessive. Slow tritium release will result in a large tritium inventory, which is unacceptable from both economic and safety viewpoints. As a consequence, considerable effort has been devoted to understanding the tritium release mechanism from ceramic breeders and beryllium neutron multiplier materials, through theoretical, laboratory and in-reactor studies. This information is being applied to the development of models for predicting the tritium release for various blanket operating conditions.

2.4.1. Theoretical studies The solid state defect structure of the lithium ceramic

(lithium vacancy, defects, traps, etc.) can strongly affect the tritium transport and release process. The origin of the lithium vacancy can arise from the following

(1) the 6Li (n,cx)3T reaction, which generates many defects in transforming 6Li into 3T and 4He atoms;

(2) defects created by displacement damage, i.e. fast neutron scattering and recoil of energetic 3T and 4He atoms;

(3) the extrinsic impurity-induced defects that control lithium diffusion;

(4) the intrinsic defects resulting from thermal equi- librium.

In in-reactor and laboratory studies, tritium release is found to be controlled primarily by two processes: bulk diffusion in the grains and desorption from the grain surface. Also, hydrogen in the helium purge gas is known to promote the release of tritium from candidate ceramic breeder materials.

To understand better the role of hydrogen in the release process, calculations have been initiated to sim- ulate directly the processes through which hydrogen can interact with lithium oxide surfaces. The methods that have been employed include a combination of ab initio techniques in crystalline and cluster environments. The method uses no adjustable parameters and may be described more accurately as computer experiments rather than as calculations. Initially, attention is being given to examination of the interaction of the hydrogen molecule with the lithium oxide surface [50,51]. The nature of the surface sites that may participate in the adsorption process are described in terms of terrace- ledge-kink (TLK) terminology. Terrace sites are asso- ciated with the regular planar locations on the flat atomic surfaces, and kinks with the corners. The terrace sites have been examined and found to be energetically unfavorable with respect to hydrogen adsorption, because of their relatively high coordination and, there-

fore, the small number of unsaturated bonds surround- ing the sites. Early simulations suggest that hydrogen undergoes dissociative chemisorption to low coordination sites.

The concepts outlined above have been developed into computer models [52-58] that include the phenomena of bulk diffusion and desorption from the grain surface [52], bulk and grain boundary diffusion, desorption and solubility [53-57], and tritium transport through open porosity [58]. Surface heterogeneity was modelled using surface sites with different activation energies for tritium desorption [59]. Some success has been achieved in modelling tritium transport and release, by assuming only diffusion and desorption processes [52,591.

The desorption activation energy may change with surface coverage, because of the existence of multiple sites for adsorption [59]. For large surface coverages, both low and high energy sites will be occupied. How- ever, it is not necessary to have different sites of adsorption for the desorption activation energy to be surface coverage dependent. The measured desorption activation energy may result from interactions between molecules adsorbed on the surface. The interaction between the adsorbed hydrogen species (OH- or H-) will affect the binding energy to the surface and, therefore, the desorption activation energy. While the computer model still needs further refinement, it can predict the tritium inventory for end-of-life experiments. How- ever, as yet, it cannot accurately predict tritium release curves from in-pile experiments over the length of an experiment.

2.4.2. Laboratory studies Tritium release experiments have been performed in

various laboratories for Li20 [60,61]. LiA102 [62], Li4SiO 4 [63], Li2ZrO3 [64,65] and LizTiO 3 [66,67]. These experiments have focused on determining the tritium extraction parameters, identifying the chemical form of the realeased tritium, and characterizing the rate-limiting process. For some materials, the tritium release rates have shown significant variance under test conditions. The reason for these variances is not fully understood but may involve poorly controlled experimental conditions, different sample characteristics or unknown mechanistic effects. What has been understood is that surfaces play an extremely important role in the tritium release process. Studies [61] have demonstrated that limiting mechanisms are very dependent upon the grain size, in that desorption is limiting for small-grained materials (under 200 gm in diameter) and diffusion is limiting for large-grained materials (over

N. Roux et al./ Fusion Engineering and Design 27 (1995) 154-166 159

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Fig. l. Tritium residence times vs. reciprocal temperature for ceramics considered: LiA102 with grain size 0.5 lam and 75% theoretic density; Li2ZiO3 with grain size 1 lam and 80% theoretic density; Li20 with grain size 16 I~m and 80% theoretic density; Li4SiO4 with grain size 23 Ixm and 90% theoretic density.

2000 gm in diameter). Desorption has been determined to be the rate-limiting step in several cases [62,68].

Adsorption desorption experiments of water or hydrogen on lithium ceramics have examined the surface release mechanisms. Experimental data on water desorption from LiA102 and Li4SiO 4 have been analyzed [69], and it was concluded that the process involved multiple desorption sites exhibiting different activation energies. The desorption of water from LiA102 has been studied and sites with different activation energies have been identified [70,71]. Studies [72-74] using Fourier transform IR spectroscopy have reported several different OH (OD) species on a lithium oxide surface, and have identified surface exchange reactions using hydrogen isotopes. HT and HTO were the chemical forms of the desorbed tritium. The reaction rate for HT release was proportional to the residual tritium concentration and to the square root of the hydrogen concentration. This implies that the dissociation of hydrogen occurs on the ceramic surface, which is in agreement with the above theoretical calculations.

2.4.3. In-reactor experiments In-reactor experiments combine the effects of neu-

trons, temperature and purge gas chemistry in a single test. To date, tritium release experiments have been conducted for the candidate materials: Li20 [75,76], LiA102 [75,77], Li2ZrO3 [75,77-79] and Li4SiO4

[77,80]. One of the capsules in the BEATRIX II experiment [76] contained an Li20 ceramic pellet stack expe- riencing a large temperature gradient, and provided some engineering performance data. In the in situ tritium release experiments, tritium residence times of 1 day were found in helium with 0.1% H2 purge gas at about 300-320 °C for Li20 and LizZrO3, 390 °C for Li4SiO4, and about 450 °C for LiAIO 2 (see, for example, Fig. 1 which displays tritium residence times vs. the reciprocal temperature for candidate ceramic breeders tested in Siloe under comparable conditions [75,80]). This is extremely satisfactory. The chemical form of the released tritium was HT and HTO, the fraction of each depending on the oxygen potential of the system.

The tritium diffusivities in the grain were measured for some of the ceramic breeders. The tritium diffusivity in the grain boundary was also estimated [81,82]. In most experiments, the tritium release was not fully explained by bulk diffusion. Experiments [83,84] have shown that desorption is a first-order reaction, and the activation energy for desorption depends on the surface coverage. The surface reaction was also studied by varying the H 2 pressure. A one-half power dependence of the surface desorption rate constant for single-crystal Li20 , and unity power dependence for sintered Li20 pellets have been reported [85]. The tritium inventories were measured at the end-of-life for several experiments and were found to be well within current safety design criteria.

Current ceramic breeder blanket designs include beryllium for neutron multiplication. Small quantities of tritium are generated in the beryllium via a 9Be neutron reaction producing tritium and helium. Be- cause the build-up of the tritium could become a safety issue, there is strong interest in tritium removal. Initial tritium release studies on beryllium have been on material that has served as a reflector in a nuclear reactor. Generally, these reflector materials have remained in place for several years at low temperatures, e.g. under 100 °C. When these reflector materials have been examined under isothermal annealing conditions, they have exhibited burst releases of tritium at high temperatures. Experiments on lower density beryllium show significant tritium release at lower temperatures.

Annealing of high density (about 100% of theoretic density) beryllium discs from the SIBELIUS experiment [48,49,86] have shown that the bulk of the generated tritium was retained in the beryllium and, when the discs were heated to 650 °C and above, the tritium was readily released. The results indicate that tritium release from the beryllium did not exhibit a burst release behavior, as previously reported, but rather showed an


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orderly release dependent solely upon the temperature. Generally, about 99% of the tritium was released by 850 °C. In comparison with literature information on tritium release from about 100% dense beryllium, the SIBELIUS data for tritium release vs. temperature show a steeper slope. Tritium release from the ceramic discs in the SIBELIUS experiment was quite similar to the behavior shown in other dynamic tritium release experiments on lithium ceramics.

2.5. Irradiation behavior

The components of an operating fusion blanket must tolerate high levels of neutron exposure. Breeder blanket lithium ceramics have the added burden of remain- ing functional while 10%-25% of the lithium atoms are transformed into helium and tritium atoms. With no significant amount of irradiation data available a decade ago, worst-case scenarios for the impact of irradiation damage in lithium ceramics were generated to provide a theoretical envelope for designers to work with. For example, the temperature limits for lithium oxide were considered very restrictive at one time [87]. Uncertainty in the irradiation performance and the volatility of lithium were considerd to be limiting issues. In addition, it was considered possible that irradiation effects in lithium ceramics could significantly increase the tritium inventory at high burn-ups. Today, irradia-

tion test results indicate that candidate lithium ceramics readily survive the rigors of neutron irradiation and achieve significant burn-ups.

2.5.1. Materials performance Numerous tritium release experiments have been con-

ducted at low lithium burn-up levels to identify the tritium release kinetics. The functional requirement of the blanket is to achieve a very low tritium inventory during operation and to minimize releases during an accident scenario. Results from the BEATRIX II experiments have provided in situ tritium recovery data at low and extended burn-ups (5%) for lithium oxide and lithium zirconate. The data in Fig. 2 show a low, stable tritium inventory at a high burn-up level. Although the shape of the curve changed somewhat during testing, the inventory changes (i.e. 0.14, 0.06, 0.09 Ci) confirm that no increase in tritium inventory resulted from irradiation damage [88]. For ternary ceramicS, i.e. LizZrO3, LiA102 and Li4SiO4, the effects of irradiation to 3% burn-up in a thermal neutron environment produced very little change in the tritium release behavior [77]. The ongoing irradiation of lithium ceramics to 10% bum-up should contribute to our confidence in maintaining a low tritium inventory at operating temperatures.

As part of the investigation of lithium ceramics' behavior in a neutron environment, short-term irradia-

N. Roux et al. /Fusion Engineering and Design 27 (1995) 154-166 161

tions were conducted on the ternary ceramics in thermal neutron reactors. They are characterized by low lithium burn-ups (1%-3%) and low displacement per atom (dpa) values (about 2), temperature levels of 400, 600 and 800 °C, and high heating rates, inducing large thermal gradients. Post-irradiation examination focused on the dimensional stability, physical integrity, and microstructural and mechanical testing [33,89,90]. Fragmentation, which was assigned to result mostly from large thermal gradients prevailing in the specimens and from thermal shocks resulting from reactor trips, was observed in some cases. Decreases in the Young's modulus and fracture strength values could be noted, which are related to the observed cracking. The thermal conductivity, as estimated from records of ther- mocouples inserted in the center and edges of the specimens, was found to be very stable. On the whole, no significant changes were observed under the conditions explored.

Data from fast neutron reactor experiments are more relevant because they correspond to higher burn-ups and greater displacement damage. The ternary ceramics were found to possess relatively small swelling rates while Li20 exhibited high swelling at modest burn-up levels (3%). However, radiographs of the FUBR 1 B (second insertion) experiment [91], with burn-ups in the range of 8%-10% of lithium, demonstrated the stability of the materials under very severe temperature gradients resulting from the large pellet diameter and high heat generation rates. For example, LiA102 was exposed, for almost 1 year, to a centerline temperature of 1125 °C, while Li2ZrO 3 was subjected to a centerline temperature of 1225 °C with an edge temperature of 600 °C. From the radiographs, it was noted that the fine-grained LiA102 pellets actually contracted, while the larger grained LiA102 pellets expanded slightly. Considering the potential for sintering of fine-grained ceramics in this extreme environment, these results are consistent. The Li2ZrO3 pellets appeared to expand slightly in both the axial and radial directions, though perhaps more from crack opening displacements than from microscopic swelling. The Li4Si04 and Li20 pellets appeared to expand until the gap around them was closed. Both these materials are considered to be relatively plastic under these conditions, so no deformation of the cladding was observed. Smaller samples under smaller temperature gradients exhibited similar behav- iors. The swelling in the Li20 pellets appears to be the same as that observed at lower burn-up levels.

A detailed examination of the radiographs reveals that both axial and radial cracks existed as a result of the extreme thermal stresses. The cracking may be the

source of the swelling observed in the ternary ceramics. However, in the case of the fine-grained LiA102 column, the thin disks at the top of the column did not exhibit significant cracking. The cracking in the pellets resulting from the large thermal gradients can be allevi- ated by appropriate designing, such as thin-walled rings or small-diameter spheres [1,2]. In this experiment, cracking of spheres was not observed and cracking was not prevalent in the smaller diameter, high burn-up ternary ceramics which possessed less than 100 °C of a temperature difference across the pellet.

For Li20, in particular, the transport of lithium was considered to be a design-limiting phenomenon. A design requirement is that the breeder material should not be relocated such that neutron streaming or decreases in the designed tritium breeding ratio might occur. In the BEATRIX II experiment, the operating temperatures in the centerline of the solid pellet were measured to be 1000 °C with an edge temperature of 550 °C. Quantitatively, downstream transport of lithium was not observed in piping or other components. A central hollow core developed in these pellets, but that was considered to be the result of sintering rather than vapor transport [92]. Recent thermodynamic data [93] support the experimental finding that little lithium transport is to be expected, because the vapor pressure of lithium hydroxide--the species transporting lithium--is a factor of 30 lower than that previously stated. Hence, lithium transport does not appear to be an issue, even for Li2 O, under these extreme conditions.

Concern was expressed that neutron irradiation may degrade the thermal conductivity of lithium ceramics and, therefore, limit the lifetimes of these materials. After irradiation to 3% burn-up, there was a significant reduction in the thermal conductivity of Li20 and LiA102 at low temperatures, i.e. under 400 °C. At higher temperatures, the thermal conductivity values were unchanged or were even higher than the unirradi- ated values [94]. These results were confirmed by in situ measurements in the BEATRIX II experiment [88], because the centerline temperatures of the Li20 specimen during irradiation were equivalent to those pre- dicted from literature data and were seen to be stable. A similar stability was observed during irradiation of Li2ZrO 3 [79].

2.5.2. Fundamental aspects In parallel to irradiation performance testing, basic

studies were initiated. Irradiation-induced defects, such as F ÷ centers, colloidal lithium metal, silicon metal, and some decomposition products in Li20, LiA102, Li4SiO4 and Li2ZrO3, have been studied using neutron, heavy-


ion, or electron irradiation to gain a fundamental understanding of irradiation behavior [95-100]. Funda- mental studies o f irradiation effects on lithium ions have been carried out for Li20 and Li4SiO4, by mea- suring the ionic conductivity of ion-irradiated specimens [101,102]. These low temperature (under 350 °C) studies may relate to tritium transport in lithium ceramics, assuming that tritium diffusion in lithium ceramics is closely related to the lithium ion diffusion. In one study, irradiation damage in Li20 from fast neutron irradiation was investigated in a series of experiments in the BEATRIX II program [100]. Electron spin resonance and optical absorption measurements were carried o u t on Li20 after irradiation in F F T F for 300 effective full power days (fluence of 3.9 x 1026 n m -2) at about 375 °C. The electron spin resonance data were interpreted to indicate the presence of colloidal lithium. Since Li20 is a very stable material, a regime of non- stoichiometry is needed to accommodate the presence of colloidal lithium at this temperature. This subject needs further study.

2.6. Activation and 6Li management

Evaluation of the neutron activation of candidate ceramic breeder materials is important for fusion reactor maintenance, waste disposal and recovery of 6Li. This evaluation depends, to some extent, on the blanket design conditions, i.e. materials proportions, neutron fluxes and neutron wall loading. Activation calculations were made for candidate ceramic breeders and structural materials under comparable conditions, allowing for a ranking of the materials. For example, the neutron activation of the materials has been calculated for a helium-cooled ceramic breeder blanket design exposed to a total integrated neutron dose of 20 MWA m -2 (5 MW m -2 and 4 years of operation) [103]. The activation levels at all times after shutdown, the biological dose rates, the decay heat values and the waste disposal ratings are reported. Likewise, calculations were made for an outboard blanket region zone of the EEF reactor with a neutron flux of 4.18 x 10 ~8 m -2 s -~, integrated dose of 12.5 MWy m -2 (5 MW m -2 and 2.5 years of operation). The specific activity at various cooling times, the surface dose rate, the ingestion and inhala- tion dose, the decay power and waste management options were considered [104]. Though conditions in the two studies are not readily comparable, the pre- dicted activities agree well within a factor of 6. The main conclusions are as follows.

(1) The dominant activity in the ceramics arises from the tritium generated. However, the level of residual

tritium is expected to be very low in the ceramics after service, otherwise the burnt material will be accessible to a wide range of processes to remove tritium to the lowest practical level.

(2) In the absence of tritium and of troublesome impurities, the intrinsic activities which are dominated by short-lived nuclides are ranked as L i 2 0 < Li4SiO 4 < LiA10 2 < Li2ZrO3, with that of Li20 being three orders of magnitude lower than that of Li2ZrO3.

(3) In all cases, activation of the structural materials is higher than that of the ceramic breeder materials. Thus, the ceramic breeder material will not be the dominant contributor to the gamma activation leading to the limitation of maintenance.

With respect to waste disposal, one must consider the production of long-life radionuclides. For the ceramics, the radioactivity after 1 year is very low in comparison with that of the structural material. The long-lived radionuclides, such as 26A1 (half-life of 7.2 x 105 years) and 94Nb (half-life of 2 x 104 years), are decay products from LiA10> Li4SiO 4 and Li2ZrO3, Li20 does not present a problem for waste disposal, while careful attention to blanket design for the other ceramics should afford some relief on this issue.

Reprocessing of ceramic breeders for the recovery of 6Li was investigated for the oxide, aluminate, orthosilicate and zirconate. Dissolution of these ceramics using water and various acids, and lithium recovery processes, with precipitation methods using ammonium carbonate or ammonium hydroxide and solvent extraction methods were examined [105]. A high dissolution rate (92%-100%) and lithium recovery rate (80%-100%) were attained. Also, prelim- inary lithium recovery experiments using neutron- irradiated LiA102 were carried out by the precipitation method and it was con-firmed that lithium can be recovered with a high efficiency (about 78%). Thus, reprocessing for the recovery of lithium was demonstrated by use of this chemical process. Methods for the recovery of 6Li were also investigated in Ref. [1061.

Solvent extraction and ion exchange were considered as potential routes for the separation of lithium from the used breeder, Methods were found to dissolve the titanate [67] but lithium recovery processes have not yet been studied.

To evaluate the feasibility of reprocessing of ceramic breeder materials, one must consider radioactivity and residual tritium. The residual tritium levels for ceramic breeder materials may not be a problem, since the solubility of tritium in the ceramic breeder material is very low and most of the tritium can be released from

N. Roux et al. / Fusion Engineering and Design 27 (1995) 154 166 163

the ceramic by heating before reprocessing. With respect to gamma activation, consideration may not be needed for Li20 and Li4SiO4, while some consideration must be given to LiA102 and Li2ZrO 3. However, it is shown that all materials can be handled remotely in radiologically controlled enclosures with a moderate degree of gamma-ray shielding within a cooling time of 10 years [106].

3. Conclusions

This summary of experimental results shows that aspects relevant to the utilization of ceramic breeder materials for fusion reactor blankets are exhaustively covered. Examination of the properties database, of out-of-reactor tests, irradiation experiments and performance analysis results highlights the functionality of ceramic breeder materials and does not reveal any negative aspects. As a result of continuous progress, current generation ceramic materials are adequate to fulfil the requirements for several attractive ceramic breeder blanket options being developed for both near- term machines and for power reactors. Advanced ceramic breeder materials, including other ceramic compounds (lithium titanate) as well as added value products are still expected.

Work still remains to be done. Specific areas for further research include the following: fundamental studies to understand fully and control pertinent phenomena, and help materials improvement; tritium release modelling to predict tritium inventories in the whole range of fabrication, operation and accident conditions; development of advanced materials and industrial fabrication capacity; most importantly, evaluation of materials behavior at end-of-life, i.e. 10%-30% lithium burn-up and high displacement damage.

To date, safety characteristics, a satisfactory properties database for candidate materials, successful development of pilot-scale fabrication using well-proven technology, excellent tritium release behavior, very good irradiation performance to extended burn-ups, good prospects regarding 6Li recovery, materials reprocessing and waste disposal designate lithium ceramics as excellent, first-option tritium breeding materials.

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summary of experimental results for ceramic breeder materials

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