The HICU PIE results of EU ceramic breeder pebbles: General characterization

Journal of Nuclear Materials 531 (2020) 152023

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The HICU PIE results of EU ceramic breeder pebbles: General characterization M.H.H. Kolb a, J.M. Heuser a, *, R. Rolli a, H.-C. Schneider a, R. Knitter a, M. Zmitko b a b

Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM), 70621, Karlsruhe, Germany Fusion for Energy (F4E), Josep Pla 2, Barcelona, Spain

h i g h l i g h t s
HICU PIE of neutron irradiated Li4SiO4-based ceramic breeder pebbles.
The material properties significantly depend on the irradiation temperature.
An enrichment in 6Li does not deteriorate the pebbles’ properties.

a r t i c l e i n f o

a b s t r a c t

Article history: Received 15 August 2019 Received in revised form 22 November 2019 Accepted 20 January 2020 Available online 27 January 2020

The HICU (High neutron fluence Irradiation of pebble staCks for fUsion) experiment was performed in the High Flux Reactor (HFR) in Petten, NL, in order to irradiate candidate tritium breeder materials in a fusion relevant environment. The presented work focuses on the post-irradiation examination of the irradiated lithium orthosilicate based breeder pebbles. The pebble samples showed three different contents of Li-6 and were irradiated at two different temperatures and in mechanically constrained and unconstrained state. In this particular publication, the influences of the irradiation conditions on the pebble morphology, microstructure, porosity, and mechanical strength are addressed. The results indicate that in general a high irradiation temperature seems to be advantageous for maintaining the mechanical strength of the irradiated pebbles. A higher mechanical strength and a significantly lower closer porosity is observed for samples that were irradiated at high temperatures in comparison to pebbles that were irradiated at low temperatures. The effects on the pebble properties with respect to the Li-6 content are small in contrast to effects of the irradiation temperature. With an increased Li-6 content, no deterioration of the material properties was observed, especially for samples irradiated at high temperatures. © 2020 Elsevier B.V. All rights reserved.

1. Introduction For the development of the European Helium Cooled Pebble Bed (HCPB) blanket concept, the HICU experiment was performed in the HFR in Petten, NL [1]. Since such an irradiation experiment with particular focus on breeder materials is one of the very few that was performed within the last decade, its results are highly anticipated for judging the aptness of the current grades of breeder ceramics. Furthermore, the results are expected to expose areas of lacking performance that need considerable improvement before these materials can be applied in a fusion blanket. The actual irradiation

* Corresponding author. E-mail address: [email protected] (J.M. Heuser). https://doi.org/10.1016/j.jnucmat.2020.152023 0022-3115/© 2020 Elsevier B.V. All rights reserved.

campaign spanned from February 2008 to December 2010. DEMO relevant conditions were chosen to test the behavior of several tritium breeding materials. To address a broad spectrum of irradiation conditions, the high fluence neutron irradiation was performed under different conditions with respect to the temperature and Li-6-content of the ceramic breeder (CB) pebbles up to damages of 20e25 dpa (in steel). To reduce the thermal neutron flux, and thus to be more DEMO relevant, the sample holders were shielded with cadmium. For the irradiation, the CB pebbles were welded in EUROFER steel capsules that were lined with platinum foils to prevent direct contact and therefore corrosion of the steel capsules and the pebble samples. Four types of lithium orthosilicate (Li4SiO4) based pebbles supplied by KIT, Germany, were irradiated in this campaign. These were fabricated with a surplus of 2.5 wt% silica (OSi) at SCHOTT AG,

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Germany. These grades represent the former EU reference CB material for the ITER Test Blanket Module (TBM). During the irradiation high and low temperature conditions were realized. Furthermore, the pebble beds were irradiated in pre-strained and in unstrained conditions. These samples were split for the various irradiation conditions, so that in total 10 different, irradiated samples are available. To be precise, in total 12 capsules that contain lithium orthosilicate based breeder pebbles were retrieved from the irradiation rig. Of these 12 capsules, two pairs were irradiated under the same conditions. Therefore, with respect to the irradiation conditions, 10 different samples are available. After the irradiation, the samples were stored at NRG and, after an initial disassembly of the irradiation rig, transported to KIT’s Fusion Materials Laboratory for the Post Irradiation Examination (PIE) in 2015. A short summary of the HICU PIE results was already included in an overview article on functional materials for the EU HCPB TBM [2], a detailed presentation and discussion of the results are now published in two articles. In this publication, the general characterization of the irradiated samples by post-irradiation examination (PIE) is presented and discussed, while the tritium release from these samples is discussed in an additional publication [3]. The general characterization includes the investigation of the pebble morphology and their porosity and their mechanical strength. 2. Experimental In the following, the examined samples are listed and the applied methods of characterization are detailed for each sample. As already stated, the HICU irradiation campaign comprised a number of different irradiation conditions. Most importantly, the irradiated samples differed in their initial Li-6 contents:
Natural abundance: 7.5 % Li-6
Li-6-depleted: 0.06 % Li-6
Li-6-enriched: 20 % Li-6 Furthermore, two different irradiation temperatures were realized:
Low temperature (LT): about 650  C
High temperature (HT): 800e850  C Additionally, the samples were irradiated in pebble beds of two different states of mechanical strain:
No initially applied stress (pebble beds of low constraint (LC))
Uniaxially pre-compacted pebble beds at 0.7 MPa (highly constrained pebble beds (HC)) All but one sample were fabricated by the lithium carbonate or lithium orthosilicate based Schott process [4]. The other sample was fabricated using lithium hydroxide (LiOH, LOH) instead of lithium carbonate and lithium orthosilicate as raw material [5]. All lithium orthosilicate pebbles were irradiated in the as-prepared state, i. e. without prior thermal annealing or conditioning. The realized combinations of these irradiation conditions are detailed in Table 1. With regard to the Li-6-burn-up, a brief estimate for a sample with natural Li-6-concentration, which witnessed the least neutron flux during the experiment, amounted to 0.6 % [6]. For a drum that kept a Li-6-enriched sample near the maximum neutron flux that was achieved in the HICU irradiation, the burn-up was calculated to 3.8 % [7]. Yet, there is no additional or more elaborate information of the amount of transmutated lithium available. After receiving the transport cask, all inner transport devices were decontaminated

and assorted. The steel drums were identified and opened by cutting. The specimens were optically inspected in a first step. For a number of capsules, some force was necessary to overcome the adherence of the pebbles between themselves and between the capsules in which they were irradiated. The removed mass of pebbles was less than the initially introduced amount for all samples. In general, a considerable amount of pebbles was found to be fragmented after retrieval. Yet, it is difficult to judge to which degree the application of force during the removing caused the fragmentation. To separate the majority of fragments from the intact pebbles, all particles smaller than 0.1 mm were removed by sieving. The remaining sample amount is considered as usable material for the characterization. For the PIE of the retrieved samples, several investigation techniques were successfully applied and are listed in the following:
Visual inspection and geometrical characterization by optical microscopy (OM)
Inspection of morphology and microstructure by scanning electron microscopy (SEM)
Mechanical characterization by uniaxial crush load determination
Determination of the closed porosity by helium pycnometry
Determination of the total porosity by Archimedean immersion All performed characterization techniques require and oftentimes also consume a certain amount of material. In some cases, the sample amount was too limited so that the determination of the total porosity could not be carried out for the following samples:
(HT, LC, 7.5)
(LT, LC, 7.5)
(HT, LC, 20) The OM was carried out by an optical inverted Olympus GX51 Light Optical Microscope which is installed in a hot cell. Bright field and dark field conditions were applied as contrasting techniques as well as polarized light methods. Controlling and image analyzing is given by software (Olympus analysis 5.1, newest update). Images are acquired by a high resolution CCD camera, which is mounted on the microscope. The dimensions, i. e. the size distributions, of the irradiated lithium ceramic pebbles were measured in this way from a loose monolayer of pebbles. For the investigation of the microstructure, a subsample of the pebbles was embedded in epoxy resin, ground and polished with water-free benzene to avoid an etching of the pebble surfaces. The pebbles were ground to almost their equator during the preparation. SEM investigations were carried out with a 40 kV RemX/Cam Scan 44. By SEM, equipped with an energy dispersive X-ray spectroscopy (EDS) detector; cross-sections of pebbles as well as the surface of a number of whole pebbles were investigated. The crosssections of the samples, which were previously used for the characterization by OM, were etched with water before the analysis in the SEM to reveal the microstructure of the pebbles. In order to mitigate electric charges on the surface of the cross-sections during the analysis, a thin layer of platinum was sputtered on the surface. For the investigation of the surface of the pebble samples, a small number of pebbles are placed on a sample holder which is wetted with conductive silver paste. Also these samples were coated with a thin layer of platinum by sputtering to mitigate charging effects. To determine the mechanical strength of individual pebbles of the HICU samples, a uniaxial compression until failure of pebbles of uniform size was performed. The maximum load that each pebble

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Table 1 List of characterized HICU samples and their irradiation conditions as well as their basic properties. Raw materials

Irradiation temperature/ C

Li4SiO4þLi2CO3þSiO2 650 800e850 Li4SiO4þSiO2 650 650 800e850 800e850 800e850 LiOH þ SiO2 Li4SiO4þLi2CO3þSiO2 650 800e850 800e850

Constraint Li-6-content/ Estimated damage in steel/ Irradiated sample % dpa mass/g

Year of production

Sample name/ abbreviation

Low Low High Low Low High High Low Low High

2002

(LT, LC, 0.06) (HT, LC, 0.06) (LT, HC, 7.5) (LT, LC, 7.5) (HT, LC, 7.5) (HT, HC, 7.5) (HT, HC, 7.5, ex LOH) (LT, LC, 20) (HT, LC, 20) (HT, HC, 20)

0.06 7.5

20

10.1 11.9 5.2 11.3 10.1 12.4 12.1 12.4 14.2 11.3

withstood was recorded as so-called crush load. In total 40 pebbles were investigated per HICU sample. For being consistent with past investigations of the non-irradiated pebble material, a pebble size of 500 mm was selected. However, within the subsamples that were used for these experiments, it was not always possible to find 40 pebbles of 500 mm diameter. In these cases, pebbles of about 500mm diameter were used instead. The pebbles are placed between two plates of BK7 glass for the compression tests. The pebble size was measured before the actual crush test as the distance between the compression plates to accommodate for these deviations. Prior to these investigations, the pebbles were heated up to 300  C for 1 h in a nitrogen environment to remove any absorbed moisture, which could reduce the strength of the samples. The used testing instrument (Zwick Precision Line) is especially qualified for automated testing of small components. A complete compensation for the machine’s compliance is integrated in the testing software (Zwick TestXpert). The used load cell allows for a maximum load of 50 N and features a resolution of 0.04 N. The whole testing instrument is integrated in a glove box system equipped with inert gas atmosphere (N2). From these crush load measurements, the mechanical stability is derived. Yet, the differences in the measurement equipment, since the determination of the crush load in the as-fabricated state 20 years ago are significant. Furthermore, the porosity of the samples is considerably different with regard to the as-prepared state of the pebbles and the irradiated state (see section 3.3). This also affects the mechanical properties of the pebbles. Especially the Young’s modulus of the pebbles is directly correlated to the pebbles’ porosity. Because the deformation of the pebbles during uniaxial loading is governed by the bulk material, rather than the surface of the pebbles, only the closed porosity of the samples is considered for the derivation of the Young’s modulus. Nevertheless, the open porosity can be a factor in the failure behavior. The following correlation of the Young’s modulus E and the porosity p was established for pure Li4SiO4 [8,9]:


 E ¼ 110  ð1  pÞ3 1  2:5  104 ðw  20Þ where w denominates the temperature in degree Celsius. It is used in the following, although the mechanically slightly stronger Li2SiO3, which, in all likelihood, is featured in the samples to a significant degree, is not accounted for by this formula. Even if a reasonable estimation of the effects of the Li2SiO3 content on the mechanical properties was used, there is an enormous uncertainty of the actual lithium metasilicate content in the sample because of the ambiguous burn-up of lithium during the irradiation. For the sake of simplicity, the temperature dependence of the Young’s modulus is also neglected. Also the published Poisson’s ratio for pure Li4SiO4 (v ¼ 0.24, [8]) is used during the following calculations, neglecting any effects of Li2SiO3 on this value.

0.70 0.69 1.86 0.69 0.70 1.87 1.85 0.69 0.70 1.80

1997

2003 2000

Experimentally, only the mechanical load on the individual pebbles can be obtained. Yet, as the mechanical properties of the plate materials are known (see Table 2), the mean contact pressure pm that acts on the elastically deformed contact region can be derived as follows [10]:

2 pm ¼ 3

sffiffiffiffiffiffiffiffiffiffiffiffi 3 6FE 2 p3 R2

with F being the applied load and R being the pebble radius and

1 1  n21 1  n22 ¼ þ E E1 E2 where the subscripts 1 and 2 denominate the pebble materials and the plate material, respectively. This concept also compensates for a differing pebble diameter, which is required for samples with little amounts of appropriate pebbles. As a result, the so-obtained mean contact pressure is independent of misleading material changes and is regarded to be a far better assessment of the mechanical stability of the pebbles. Furthermore, all peak stresses within a pebble under a compressive load linearly scale with the contact pressure [13]. In addition to the standard characterization of the samples, a Weibull analysis of the crush load data was performed to evaluate the statistics of the failure behavior of the pebbles. It also provides insight into the failure probability of a sample as a function of the applied load. A two parameter Weibull distribution was assumed for all samples, as the case for a three parameter Weibull distribution can rarely be made [14]. In this study, the mean contact pressure at which the pebbles fail is used instead of failure stress, which is usually applied. Technically, the Weibull analysis, i. e. the determination of the Weibull modulus m as well as the characteristic mean contact pressure pm,0, were determined by a maximum likelihood estimation (MLE). In contrast to the fitting of a straight line to the data points by a least-squares algorithm and thus deriving the Weibull parameters, the MLE approach directly estimates the two Weibull parameters, while the (intrinsic) weighing of the individual data points is prevented. For removing the bias from the determined

Table 2 Mechanical properties of the plate materials that were used during the uniaxial compression tests [11,12]. With the exception of the samples (non-irr, 7.5, ex LOH) and (non-irr, 0.06) which were tested with sapphire plates, all other samples were tested with the BK7 glass as plate material.

Sapphire BK7 glass

Young’s modulus/GPa

Poisson’s ratio

345 82

0.29 0.206

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parameters, the proposed functions of Hirose and Ross were used [15,16]. The confidence boundaries were determined for a confidence level of 90 % according to the likelihood ratio criterion. The exact procedure is described extensively in literature [14,17,18] and implemented in R (version 3.3.2, x64). The two dimensional confidence intervals on both Weibull parameters, as determined by the likelihood ratio method, can be visualized in the form of contour plots [17]. When two or more datasets are to be compared, these likelihood contours can also be an indicator whether the datasets are statistically inequivalent within the given confidence level. If two of these shapes do not intersect, the failure behavior of the two sample sets is dissimilar with 90 % confidence. However, the opposite is not necessarily true. The density of the HICU pebbles was determined by helium pycnometry. The quotient of the so-measured density and the theoretical density of the material is interpreted as the closed porosity of the samples since helium penetrates, and thus disregards, all pores with access to the atmosphere (so-called open pores). The theoretical density which is used in these calculations takes neither lithium losses due to transmutation, potential evaporation or other effects into account nor changes of the phase composition that may originate from the loss of lithium, because the factual lithium losses could not be estimated in this work. The density resolution of the available helium pycnometer is estimated to be about 0.01 g/cm3. The total density of the HICU samples was determined by the Archimedean immersion method, enhanced by the use of a glass pycnometer. For this method, a subset of each sample was immersed in inert water-free organic oil. Due to the surface tension of the oil, it does not penetrate the open pores and thus, open and closed pores are included in the determined density. The total porosity is also determined as a fraction of the theoretical density (also without taking lithium losses into account). For the determination of the density, the volume of the oil with the immersed pebbles was precisely measured with a glass pycnometer. The mass of the immersed pebbles as well as the mass of the used oil was determined before. To decrease the effects of the environment on the measurement results, the temperature was kept constant during the individual measurements. The density resolution is also estimated to be of the order of 0.01 g/cm3. However, to achieve a reasonably high precision, a significant amount of pebbles must be used and therefore the measurements could not be conducted for all samples, as stated above. To investigate the effects of neutron irradiation on the pebble properties as much as possible, the available characterization results of the respective non-irradiated samples were used. These results were obtained as part of a routine investigation in a short timespan after their production by the same methods (cf., Table 1), yet using state of the art measurement devices at that time. Since the irradiation of the samples was carried out at elevated temperatures, some changes of the pebble properties and morphologies that have developed during the HICU experiment may be the result of the involved annealing of the pebbles rather than the actual neutron irradiation. To separate the annealing from the irradiation effect, also the available data of annealed samples, i. e. pebbles with natural Li-6-content that were fabricated from Li2CO3 and Li4SiO4 and were annealed for 96 days at 970  C, or the available asfabricated samples of depleted and enriched material as well as the (ex LOH) sample with natural abundance that were annealed for 7 days at 900  C, were characterized accordingly.

figures and graphs.

3.1. Visual inspection and light microscopic characterization As already mentioned in the previous section, the pebbles had to be removed from the irradiation capsules more or less forcefully. Therefore, it is not surprising that a considerable number of fractured pebbles are retrieved and that less than the initially irradiated amount is retrieved. However, the amount of retrievable material varies considerably and unsystematically between the samples with losses of up to 74% as a fraction of the initially irradiated material mass. This is similarly true for the sample mass, for which the particle size undercuts 0.1 mm. The amount of these pebble fragments material varies, also unsystematically up to 26 % as a fraction of the initially irradiated material mass. The visual inspection of the samples revealed a heterogeneous color change in the pebbles. The observed darkening is probably a result of oxygen vacancies provoked by the ionizing radiation. The color intensity differs for the several pebble samples (see Fig. 1). Samples irradiated at higher temperatures show in most cases less tendency to a color change, whereas samples irradiated at lower temperatures display dark pebbles more frequently. Also the degree of the darkening is usually more pronounced for the samples irradiated at the lower temperature. This could be an effect of

3. Results The results obtained from the different characterization techniques are summarized in this chapter and illustrated by selected

Fig. 1. Photographs of the HICU samples after retrieval. All samples show a heterogeneous darkening as a result of the irradiation. (Left: samples irradiated at low temperature; right: samples irradiated at high temperature.)

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thermal recovery of radiation induced F-centers at higher irradiation temperatures. For all samples, some broken pebbles can be identified. They may have formed during irradiation as well as during the forceful removal from the irradiation capsules. Generally, the surface of the irradiated pebbles appears to be matted, whereas the as-produced pebbles usually feature a shiny surface. The development of the microstructure analyzed by light microscopy can be seen in Figs. 2 and 3 exemplarily for the batches enriched with 20 % Li-6. The evolution of cracks and pores observed in light micrographs in the HICU samples compared to the asprepared material varies with the irradiation temperature. From these images it is furthermore evident, that all irradiated pebbles, irrespective of their irradiation conditions or Li-6-content, reveal a dark layer at the pebble surface, as shown in Fig. 4 for the (HT, LC, 7.5) sample. The thickness of the layer varies from 1 to 15 mm also within one sample. SEM micrographs of cross-sections hardly reveal this layer on the edge of the pebble. Element analyses using EDS were tested to determine the chemical composition of the phase in this surface layer. Due to the technical circumstances, an unequivocal analysis of the surface layer was not possible. Yet, an accumulated carbonate region formed by CO2 adsorption during the long storage of the irradiation rig in ambient air could be a sensible explanation. The results of the thermal desorption experiments that were carried out with these samples, see Heuser et al. [3], also support the assumption of a lithium carbonate layer at the surface of the pebbles. For all pebbles with natural Li-6 content or 20 % Li-6-content, the irradiation at higher temperatures leads to a microstructure with a few large pores and a number of cracks, while pebbles irradiated at lower temperatures reveal an increase in the number of large pores. However, an increase in large pores and also fewer cracks were already found after an annealing of the initially prepared sample (900  C, 7 days). In contrast to that, the Li-6-depleted samples show a significantly different behavior when the same comparison is performed (see Fig. 3). In this case, no pores of significant size were formed as consequence of the thermal treatment. With respect to the irradiation temperature, a higher temperature

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Fig. 3. Optical micrographs of cross-sections of 0.06 % Li-6-depleted pebbles after treatment in different conditions. The top left image shows the pebbles in the asreceived state and the top right image after annealing at 900  C for 7 days. The two bottom images show the pebbles after neutron irradiation in the HICU experiment in low constraint conditions at low temperature (left) and at high temperature (right).

Fig. 4. Detail optical micrograph of a non-etched cross-section of sample (HT, LC, 7.5), which exemplarily shows the dark layer that is observable for all samples.

Fig. 2. Optical micrographs of cross-sections of 20 % Li-6-enriched pebbles after treatment in different conditions. The top left image shows the pebbles in the asreceived state and the top right image after annealing at 900  C for 7 days. The two bottom images show the pebbles after neutron irradiation in the HICU experiment in low constraint conditions at low temperature (left) and at high temperature (right).

seemingly leads to spherical pores whereas those are not observable for the pebbles that withstood the irradiation at lower temperatures. Yet, a comprehensive analysis of light micrographs and the development of the microstructure that includes the samples with natural abundance as well as the pure thermal annealing cannot be made, as relevant microscopic pictures of the nonirradiated pebbles are only scarcely available and the original material is also not available anymore. Therefore, only these few comparisons could be considered. As mentioned above, the comparison of optical micrographs of pebbles irradiated at low temperatures with those at high temperatures exhibits in general less pores and less pronounced cracks for samples irradiated at high temperatures. Furthermore, an increased porosity is found in samples with 0.06 % Li-6-content. In Fig. 5, the microstructure of selected samples is shown with the focus on the irradiation temperature. The upper two images A and E ((LT, LC, 0.06) and (HT, LC, 0.06), respectively) compare both irradiation temperatures for Li-6-depleted, low constrained material. Both samples show a considerable number of cracks and in addition to that, the pebbles irradiated at the higher temperature also show a considerable amount of larger pores. In both cases, numerous defects are located within the volume of the pebbles. Especially for the (HT, LC, 0.06) sample, large cavities are formed in

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peculiarity, large cavities are formed within the samples similar to those that are observable in sample (HT, LC, 0.06). As a result of the applied pre-compaction, some flattening at the contact points of the pebbles is observed after the irradiation at the higher temperature. This flattening is presumably induced by moderate creep. With regard to the crack distribution within the pebbles, a clear tendency is difficult to observe. In both samples (LT, HC, 7.5) and (HT, HC, 7.5), cracks usually reach the surface of the pebbles. Seemingly, the cracks that reach the surface in sample (HT, HC, 7.5) are in tendency larger than those of sample (LT, HC, 7.5), where also larger cracks within the volume of the pebbles seem to exist. For the Li-6-enriched pebbles of the samples (LT, LC, 20) and (HT, LC, 20) also the number of large pores is significantly larger for the pebbles that were irradiated at lower temperature compared to the ones irradiated at high temperature (see Fig. 5 D and H). For the presented set of samples, the pores are the largest spherical pores compared to the other samples, of course the large cavities that are seen in the samples (LT, HC, 7.5) and (HT, LC, 0.06) are still larger. Also the pebbles of sample (HT, LC, 20) show a perceivable amount of pores, which are much smaller in comparison and lie in a similar range as those of sample (HT, HC, 7.5). Similar to the samples (LT, HC, 7.5) and (HT, HC, 7.5), these two samples do not show a clearly different microstructure in terms of cracks. Sample (HT, LC, 20) shows some large cracks that extend to the surface of the pebbles but also some cracks which formed within the volume. For the pebbles of sample (LT, LC, 20), cracks mainly seem to exist within the pebble volume, especially the large cracks. Yet, there are also some large cracks that reach the pebble surface. 3.2. Microstructure and morphology

Fig. 5. Optical micrographs of non-etched cross-sections of selected samples, which cover the whole range of the irradiation conditions.

the center of the pebbles. Yet, the cracks that are found within the pebbles of the (LT, LC, 0.06) sample do not reach the surface to the same extent as for the (HT, LC, 0.06) sample, which was irradiated at high temperatures. Also, comparing both samples, the cracks that form within the volume of the (LT, LC, 0.06) sample that do not reach the surface of the pebble, are usually larger than the cracks that form within the center of the pebbles of sample (HT, LC, 0.06). The samples (LT, LC, 7.5) and (HT, LC, 7.5) are compared within Fig. 5 B and F. Both samples feature a natural Li-6-content and were irradiated without artificial mechanical constraint. A similar behavior as for the Li-6-depleted samples can also be observed for these two samples. Yet, in this case, the formation of large pores is relatively mild and in tendency stronger for the pebbles that were irradiated at low temperature. But there is a clear trend that cracks, which do not reach the pebble surface, mainly form within the volume of pebbles that were irradiated at low temperatures, while numerous large cracks extend to the surface in pebbles that were irradiated at high temperatures. The samples (LT, HC, 7.5) and (HT, HC, 7.5) (see Fig. 5C and G) of the pre-compacted pebble beds show, with regard to the formation of large pores, the same tendency as the samples (LT, LC, 7.5) and (HT, LC, 7.5), but much more pronounced. There is a considerable formation of large pores within the volume of the pebbles that were irradiated at low temperatures in contrast to only a few, if any large pores within the pebble volume of the samples that were irradiated at high temperatures. For some pebbles, as an extreme

An overview of the development of the pebbles’ microstructure using SEM is given in Fig. 6. Exemplary for all samples, only the 20 % Li-6-enriched OSi cross-sections are depicted, which show representative microstructures for all samples. After synthesis, the pebbles show a characteristic dendritic microstructure of the main phase lithium orthosilicate (Li4SiO4) in dark gray and the minor phase lithium orthodisilicate (Li6Si2O7) in light gray. The metastable minor phase (Li6Si2O7) is decomposed to lithium ortho- and metasilicate (Li4SiO4 þ Li2SiO3) by thermal activation. Generally both of the irradiation temperatures of the HICU campaign are high enough to enable the decomposition. Therefore, the irradiated pebbles show, beside the main phase Li4SiO4 which also appears in dark gray, Li2SiO3 in light gray as the second phase. The tritium breeding reaction that takes place during neutron irradiation and the consequential loss of lithium leads to the formation of additional Li2SiO3 in the irradiated pebbles. An estimation of the amount of so-removed lithium is not possible as the neutron spectrum during irradiation is not available for this study. It must also be mentioned that the neutron spectrum varies considerably as a function of the sample position within the HICU rig and consequently, the formation of Li2SiO3 varies with the sample location too. Yet, it is always clear that with increasing Li-6-enrichment the tritium breeding reaction is significantly more probable and therefore the formation of Li2SiO3. Significant grain growth took place during the irradiation at high temperatures and some of the metasilicate accumulated at grain boundaries in the form of seams. Yet, a considerable amount of the Li2SiO3 grains remained encased within the larger Li4SiO4 grains. Between these remaining Li2SiO3 grains, thin platelets of Li2SiO3 form for all pebbles irradiated at high temperatures (see Fig. 7), also for the Li-6-depleted samples. Yet, the amount of the formed platelets seems to decrease with decreasing Li-6concentration. These platelets form in parallel to each other within a single Li4SiO4 grain. It is therefore highly likely that this

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Fig. 6. SEM micrographs of etched cross-sections of as-received and neutron irradiated samples that feature a Li-6-enrichment of 20 %. The images A, C, and E show overview images of representative pebbles, while the images B, D and F show detail views of the characteristic microstructure.

Fig. 7. Detail SEM micrograph of an etched cross-section of the (HT, HC, 20) sample, which shows the platelets that form within the Li4SiO4 grains during irradiation.

phenomenon is due to a currently unknown crystallographic relationship. As these platelets also appear within the Li-6-depleted pebbles, it is possible that these crystallographic planes act as migration paths for the excess silicon and oxygen ions at sufficiently high temperatures and that irradiation induced defects on these planes act as nuclei for the crystallization of Li2SiO3. Compared to the samples that were irradiated at high temperature, the pebbles that were irradiated at 650  C show some significant differences. The segregation of the existing and formed Li2SiO3 to the grain boundaries of the Li4SiO4 grains does not happen to a comparable extent. For these samples, Li2SiO3 predominantly remains within the Li4SiO4 grains. Some grain growth is also observable for the samples that were irradiated at low temperatures; however, that is mainly observed for the Li4SiO4 grains and to a far lesser degree for Li2SiO3. In addition to that and in support of the hypothesis that excess silicon and oxygen migrate along certain crystallographic planes, the formation of Li2SiO3 platelets within the Li4SiO4 grains was not observed. The total amount of Li2SiO3 that was formed during the irradiation, however, is seemingly less than the micrographs suggest for the samples that were irradiated at the higher temperature. Of course, this impression can be deceptive since the Li2SiO3 is more finely dispersed in the pebbles that were irradiated at 650  C. This is the result of less promoted grain growth of Li2SiO3 as in the samples that were

irradiated at high temperatures. It may, however, also be a justified impression as the samples for low temperature irradiation were located at the top and bottom of the HICU rig, where the neutron flux is significantly lower, which leads to a considerably lower heating of the pebbles, than in the center section of the rig [7]. Fig. 8 shows the surface of two representative pebble samples ((HT, LC, 7.5) and (HT, HC, 7.5 ex LOH)). It is apparent, that both samples are covered by some sort of layer. This layer could consist of pulverised particles adhering on the pebbles’ surfaces or could be identical to the dark layer that was observed in the optical micrographs (see Fig. 4) or a combination of both. As it is very likely that the layer observed in optical micrographs is a reaction layer due to the storage of the samples in air (see also section 3.1), it does not represent the actual structure of the surface that developed during the irradiation. Furthermore, the thickness of this layer and its partly very rough morphology (see the (HT, LC, 7.5) sample in Fig. 8 A) prevents a sensible investigation of the structure underneath it. Self-evidently, this layer is observed as a dark layer in the crosssecions of the samples (see Fig. 4). Since all irradiated samples show such a strong layer, the investigation of the sample surface did not provide additional information. 3.3. Porosity measurements The densities of the irradiated samples are reduced compared to those of the non-irradiated pebbles (decrease of about 7 %). Inversely, the closed porosities calculated from those density measurements result in a slight increase from 1.9 % on average for the as-prepared samples to about 2.4 % on average for the heattreated samples. A significant increase to a value of <8 %> for all

Fig. 8. SEM micrographs of the surface of representative pebbles of the (HT, LC, 7.5) and (HT, HC, 7.5 ex LOH) samples in A and B, respectively.

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samples can then be observed for samples irradiated at high temperatures and to <10 %> for all samples irradiated at low temperature. The development of the closed porosity of pebble samples that were fabricated identically is depicted in Fig. 9. The densities and the resulting closed porosities reflect the observations of the microstructural analyses. Total porosities were derived from densities measured by immersion method using oil as immersion liquid. As described in section 2, measurements could not be performed for all samples and therefore, the number of comparable results is limited. The obtained total porosities after the irradiation show similar values of <15 ± 1 %> and therefore increased significantly as expected (see Fig. 10 A). The open porosity was calculated from the total porosity and the before mentioned closed porosity. Values for open porosity vary from 5 to 8 % with mean values for HT and LT of resp. <7 %> and <5 %> (see Fig. 10 B). (Due to the limited data, each column presented in Fig. 10 represents only one data point.) As far as possible, from each batch one sample irradiated at low and one at high temperatures was measured resulting in 7 out of 10 samples. Two of three samples irradiated at low temperatures show higher open porosities compared to samples irradiated at high temperatures. The most remarkable difference can be observed for the samples enriched in Li-6. Here the open porosity exhibits an opposing behavior compared to the other samples with regard to the irradiation temperature. However, no clear trend can be examined and the limited data does not allow to draw a final conclusion. In contrast to a sintering process, where the open porosity decreases and the portion of closed porosity increases, an increase for both, the open as well as the closed porosity, can be observed for the HICU samples after irradiation (Figs. 9 and 10). This can be explained by the production of tritium and helium during the neutron irradiation. Although the initial gas inclusions in the material degassed at the elevated temperatures, a steady gas production led to a swelling of the pebble and the closed as well as the open porosity significantly increased. In general, for the evaluation of the porosity, two aspects have to be considered beside the uncertainty of the theoretical density (see section 2). On the one hand, porosities were calculated from

Fig. 9. Closed porosities of the available pebble samples as a function of their Li-6content and their treatment (as-received, annealed, and irradiated at high temperature (HT) or low temperature (LT)). The annealing conditions of the samples are given in section 2. The bars indicate the mean absolute deviation of multiple samples, when applicable.

measured densities. Furthermore, the values for open porosity are directly related to the values for closed porosity. As the total porosity values are quite similar, the differences in open porosities are dependent on the differences in closed porosities. On the other hand, it has to be considered that the pebbles were stored over several years inter alia in ambient air for some time after the irradiation. The assumed lithium carbonate layer (cf. section 3.1) formed in contact with CO2 in air can have a noticeable influence on the calculated porosity of the samples, since the assumed theoretical density is no longer valid in this case. Generally, the formation of a significant amount of Li2CO3 would lead to an overestimation of the porosity of the pebbles. (An assumed 10e15 mm layer of lithium carbonate of a 300mm pebble amounts to 10e14 vol% and would result in an overestimation in the porosity of about 1e2 %.) 3.4. Mechanical characterization Fig. 11 details the average mean contact pressures for the HICU samples as a function of the used initial material (and therefore, also as a function of Li-6-content) and as a function of the irradiation temperature in comparison to the material in the as-prepared state as well as in the annealed state. It is clear that the strength of the pebbles is reduced with annealing as well as irradiation, which of course also involves an annealing of the samples. The most significant decrease, with respect to the as-received state, can be observed for the Li-6-enriched HICU sample, which is irradiated at low temperature. In general, there seems to be a tendency for the depleted as well as for the samples with natural Li-6-content that the irradiation at low temperature leads to slightly higher degradation of the pebble strength compared to irradiation at higher temperatures. For the (ex LOH) material, the effects of annealing and irradiation on the mechanical strength are rather mild. If the same trend applies as for the samples with natural Li-6-abundance, which were fabricated from Li2CO3 and Li4SiO4, the degradation of the strength at low irradiation temperatures should be low. For these samples only a slight degradation is observed after irradiation at high temperatures compared to the annealed state. Also the reduction of the pebble strength is not too severe for the Li-6-depleted pebbles, for which no significant loss of strength was suspected because of the marginal lithium burn-up. The most significant loss of strength, compared to the as-prepared state, is found for the Li-6-enriched samples. This is not unexpected as for these pebbles, the irradiation should lead to the most significant changes, since the highest amount of lithium is burnt. Overall, the strengths of the samples differ notably but not excessively. In Figs. 12 and 13 two Weibull plots of the non-irradiated and irradiated (HT, HC, 7.5, ex LOH) samples are shown exemplarily, also with the appropriate confidence boundaries. From these plots, it is clear that a single Weibull line represents the data reasonably well, although a slight tendency to higher contact pressures than predicted by the Weibull line is visible at low failure probabilities (below 10 %). As for these samples, this phenomenon is visible for all HICU samples. The reason for this observation is not fully understood, yet potentially the forceful retrieval of the pebbles from the capsules led to the breaking of the weakest pebbles, which are thus underrepresented in the available amount of pebbles. Another reasonable explanation would be the creation of a microstructure related crack deflection as a result of the irradiation, which increases the pebble strength at low loads and leads to the same observation. An overview of the determined Weibull parameters of the samples with natural Li-6-abundance is given by the likelihood contour plot in Fig. 14. The Weibull parameters of the non-

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Fig. 10. Total porosities A and open porosities B of the available pebble samples as a function of their Li-6-content and their treatment (as-received, annealed, and irradiated at high temperature (HT) or low temperature (LT)). The annealing of the samples was carried out at 900  C for 7 days.

Fig. 11. Average mean contact pressures of the as-prepared samples and heat-treated samples compared to samples neutron irradiated at high and low temperatures. For all non-irradiated samples, pebbles of 500 mm diameter were used. The bars indicate the standard deviation of the data.

irradiated samples cluster at the lower right corner of the plot, which indicates a high characteristic mean contact pressure and a comparably low Weibull modulus. However, there is one exception with the sample that was fabricated in 1997 and was annealed for 96 days in He/H2-atmosphere. It also shows a low Weibull modulus, but also a considerably reduced characteristic mean contact pressure. This is in total contrast to the irradiated samples which generally show an increased Weibull modulus with respect to the non-irradiated samples. Furthermore, the pure annealing of the pebbles also led to a considerably reduced characteristic mean contact pressure. The parameters of the irradiated samples thus show an intermediate characteristic mean contact pressure between the as-fabricated pebbles that were produced from Li2CO3 and Li4SiO4, (non-irr, 7.5, ex LOH) samples, regardless of their annealed state, and the annealed sample, that was fabricated in 1997, at about 4 GPa. All but one set of parameters can be embraced into one region at this characteristic mean contact pressure together with a Weibull modulus of about 20. Even the likelihood

Fig. 12. Weibull plot of the non-irradiated ex LOH sample with natural Li-6-content.

contours of 3 irradiated samples that were fabricated in 1997 overlap, while the irradiated ex LOH sample behaves significantly different than those, but is still in reasonably vicinity to them. The only exception from this clustered region is sample (LT, LC, 7.5). Similarly to sample (LT, LC, 0.06) (see Fig. 15) sample (LT, LC, 7.5) was annealed at low temperatures and the pebbles were not artificially constrained. Both samples show a significantly lower characteristic mean contact pressure, but a noticeably higher Weibull modulus than their respective samples that were either irradiated in a highly constrained pebble bed or at high temperatures or both.

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Fig. 13. Weibull plot of the irradiated ex LOH sample with natural Li-6-content.

From Figs. 12 and 13 it is also obvious that the pebbles that fail at high loads, lose a considerable portion of their strength due to neutron irradiation (and also to a lesser degree due to annealing which is not shown here), while the low strength pebbles are relatively unaffected. This becomes manifested in a higher Weibull modulus and also a lower characteristic mean contact pressure and

is the result of the growth of defects, which were comparably small in the as-fabricated state. But, it is difficult to assign the failure determining defects purely to volume defects, which are usually linked to a high Weibull modulus [14]. Potentially both types of defects contribute to the observed failure behavior relatively equally. This is in principle also true for the pebble samples that were fabricated in 1997. From all of these samples, volume defects probably govern the failure behavior of sample (LT, LC, 7.5) the most. However, with the annealing of the pebbles for 96 days, a pronounced grain growth is associated, which is often correlated with an increases defect size within the pebbles. Apparently the annealing leads to an overall much larger defect size than what is seen during the irradiation. The defects that govern the failure behavior of these pebbles are unambiguously identified as surface defects because of the low Weibull modulus (see Fig. 14). Furthermore, pure annealing, as also the Weibull parameters that were determined for the ex LOH samples emphasize, does not increase the Weibull modulus as clearly as the irradiation did. The likelihood contour plot of the Li-6-depleted samples is given in Fig. 15. All three samples are clearly different from each other with at least 90 % probability, as no overlap of the confidence intervals is visible. The irradiation of Li-6-depleted pebbles at high temperatures does not lead to a significant degradation of the mechanical properties with respect to the annealed, but nonirradiated sample. Yet, a significant increase in the Weibull modulus is observed, which means that the size distribution of the defects (i. e. most likely cracks), which cause the failure of the pebbles at the observed loads, is considerably narrowed. For the respective sample this means that while small defects seemingly grow to a relatively uniform size, large defects seem to be less observable within the pebbles compared to the non-irradiated state, or the failure initiating cracks are now found within the pebble volume instead of the surface, where they were located before. This does not necessarily mean that the surface cracks disappear. They may just be insignificant for the mechanical behavior. For an irradiation at low temperature, the Weibull modulus increases more significantly and goes along with a significant decrease of the characteristic mean contact pressure. While there is some ambiguity about the microstructural cause for the

Fig. 14. Likelihood contour plot of the HICU samples that feature the natural Li-6-content compared to the respective non-irradiated material in the as-fabricated state as well as annealed state.

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Fig. 15. Likelihood contour plot of the Li-6-enriched and depleted HICU samples and the respective non-irradiated samples in their as-fabricated state and after annealing.

increase of the Weibull modulus for the sample that was irradiated at high temperatures, there is little reason to doubt that volume defects have to be considered as failure determining defects for the sample that was irradiated at low temperature. The likelihood contour plot of the Weibull parameters that were determined for the Li-6-enriched pebbles is shown in Fig. 15, too. Similarly to the other HICU samples, also these samples show an increased Weibull modulus and a decreased characteristic mean contact pressure. In this case, however, the sample that was irradiated at low temperature and without artificial mechanical constraint does not behave visibly different from the pebbles that were irradiated at high temperature and high mechanical constraint. For the enriched samples, the pebbles that were irradiated at high temperature and low constraint are behaving significantly differently. Yet, in comparison to the sample with natural Li-6-content, the pebbles of sample (HT, LC, 20) are perfectly in line with them (compare with Fig. 14). For the other two Li-6-enriched samples (HT, HC, 20) and (LT, LC, 20) no significant difference in their mechanical behavior can be stated as the likelihood contours overlap. Comparing these two samples with the as-fabricated and annealed, but non-irradiated pebbles of the same material, a similar point as for the ex LOH samples, which feature the natural Li-6-content, as detailed in Figs. 12 and 13 can be made. However, the data seems even clearer. A considerable change from strength determining surface defects to volume defects occurs. That leads to the observed high Weibull modulus with the accompanying decrease of the characteristic mean contact pressure. All samples show a slight right hand bend in the region of low strength too, which is equally attributed to an effective crack deflection by the internal structure of the pebbles that leads to a higher intrinsic strength at low loads. In general, the samples that were irradiated at high temperatures show a tendency to higher characteristic mean contact pressures than pebbles irradiated at low temperatures as shown in Fig. 16. Fig. 11 also agrees with this finding. In return, the samples that were irradiated at the lower of the two temperatures show in tendency higher Weibull moduli, which can mean that an irradiation at low temperatures leads to the formation of more severe volume defects that deteriorate the pebble strength than

irradiation at higher temperatures. 4. Discussion To discuss and interpret the different material properties of the samples the irradiation temperature, the Li-6-enrichment, as well as other parameters like the dose, the constraint (high/low), or the raw materials used for synthesis were considered. As the effects of the irradiation temperature as well as the Li-6-enrichment affect the pebble properties the strongest, both are discussed in separate sections. 4.1. Effect of irradiation at high temperature versus low temperature The significantly different irradiation temperatures of the HICU samples (at 650  C and at 800e850  C), affected the resultant pebble properties most visibly. The most obvious change that can be related to the irradiation temperature is the significant increase of the closed porosity for samples that were irradiated at low temperatures over the porosity of the samples that were irradiated at high temperatures. The formation of closed pores is usually attributed to the inadequate release of generated gaseous species, i. e. tritium or helium or both. As it is shown in an additional part of this work, a significantly higher inventory of tritium can be detected for the pebbles that were irradiated at low temperatures [3]. Strictly speaking, this finding cannot be seen as an indication for the trapping of tritium within the formed pores, as the purging of the pores from the tritium cannot be shown directly. Yet, it is the most probable explanation. It is, furthermore, self-evident that the closed pores are more likely to be formed when the diffusion of the gaseous transmutation products is slower than otherwise. The diffusion of tritium (and helium) is clearly a function of the irradiation temperature and consequently readily identifiable in the results. In addition to that, this hints at an oversaturation of the material lattice and grain boundaries with tritium at low irradiation temperatures, whereas this effect is certainly less significant at high irradiation temperatures. Yet, it must be noted that the presence of an apparent lithium carbonate rich surface layer on all samples, as a result of the storage conditions, leads to a deterioration of the

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Fig. 16. Likelihood contour plot for all HICU samples, color-coded for the irradiation temperature. (For interpretation of the references to color in this figure legend, the reader is referred to the Web version of this article.)

absolute porosity values as the assumed theoretical density of the pebbles is incorrect. But, it is considered that the surface layer is equally formed on all samples as a first approximation. Consequently, the observable trends are still valid. As mentioned in section 3.1, heterogeneous darkening of the pebbles can be observed, that seems to be more pronounced for the samples that were irradiated at low temperatures. Radiationinduced F-centers, that are probably responsible for the color change, can certainly recover at a higher rate at elevated temperatures. Therefore, this recovery would be more feasible at higher irradiation temperatures and the retrieved samples generally show lighter colors. As these radiation-induced point defects were not quantified, it is not possible to link them to changes of other properties. The determined crush load of the samples is also significantly influenced by the irradiation temperature. Pebbles that were irradiated at higher temperatures generally withstand higher mean contact pressures. A direct correlation to the determined porosity cannot be drawn, especially not to the total porosity which is seemingly unaffected by the irradiation temperature. In principle, an increased open porosity can lead to a decrease of the mechanical strength in absence of other large defects. In case of the HICU samples, the open porosity rather increases with increased irradiation temperature and is therefore irrelevant for the mechanical strength. The Weibull analysis finds that the samples, which were irradiated at low temperatures, show a higher Weibull modulus compared to the samples that were irradiated at high temperatures. This indicates that volume defects are more likely to cause fracture for these pebbles, whereas surface defects are the more likely cause for failure of pebbles that were irradiated at high temperatures. The optical micrographs of pebble cross-sections, see section 3.2, seem to correlate well with these observations, as the pebbles show less pores as well as less pronounced cracks for samples irradiated at low temperatures. Moreover, there seems to be a slight tendency for cracks reaching the surface more often in case of a high irradiation temperature. The increased open porosity may therefore still be an indication of a higher density of critical surface defects. It is not yet fully understood what causes the cracks to form within the volume of the pebbles at low irradiation temperatures. Either, the higher

irradiation temperature enables local creep and therefore dissipates external stresses as well as stresses that arise from the formation of gaseous transmutation products. Or these transmutation products reach the surface of the pebbles a lot quicker, and therefore the pebbles are less pressurized from the inside and external stresses play a subordinate role. Generally, a higher mobility of the pores and stronger healing of cracks is expected at higher irradiation temperatures, which may lead to higher pebble strength. This is underlined by a denser microstructure appearing in the optical micrographs. Exceptions are the two samples that are depleted in Li-6. Here, the pebbles that were irradiated at low temperatures show severe cracks, but less pores than the sample irradiated at high temperatures. Apart from the formation of cracks and pores, the microstructure of the pebbles also differs significantly, depending on which temperature the samples were irradiated at. Samples irradiated at high temperatures show a distinct accumulation of the second phase Li2SiO3 at the grain boundaries, which occurs to a much lower degree for samples irradiated at low temperatures. Yet, because of the uneven neutron flux, the formation of Li2SiO3 is most likely increased for the samples that were irradiated at high temperatures, and thus any effects may appear more pronounced in this case. Pebbles irradiated at low temperatures, however, show no excessive accumulation of Li2SiO3 at the grain boundaries, but in contrast a comparably fine dispersion within the Li4SiO4 grains. Mostly these Li2SiO3 precipitates are decorated by occurring pores. These observations also hint at the considerably increased diffusion within the Li4SiO4 grains at higher temperatures. 4.2. Effect of the lithium-6 enrichment The different batches used in the HICU experiment had three different Li-6-contents of 0.06 %, 7.5 %, and 20 %, respectively. Hence, the desired tritium breeding reaction during the neutron irradiation was either almost entirely suppressed or significantly increased to reach more relevant levels in dependence on the Li-6content of the samples. Therefore, a pronounced influence on the materials’ behavior was expected with regard to the different amounts of products resulting from the breeding reaction. The degree of the occurring color change of the neutron

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irradiated HICU pebbles does not show an apparent trend with regard to the Li-6-content. Possibly, the enriched samples reveal a slightly stronger tendency to a color change. However, independent from the Li-6-content and the initiated tritium breeding reaction by neutrons, all pebbles are exposed to ionizing radiation. The creation of defects is almost exclusively caused by the neutrons itself and not by possible interactions of the products resulting from the breeding reaction. It is therefore also sensible that the effects of the Li-6-enrichment are limited in this regard. In contrast to the obvious dependence of closed porosity from the irradiation temperature, a clearly observable dependence from the Li-6-enrichment cannot be substantiated. This is in reasonable accordance with the little differences that were found in the amount of released tritium at high irradiation temperatures [3]. Also the optical micrographs do not suggest the existence of any trend. Yet, for the (HT, LC, 0.06) sample, the high amount of observable porosity, which is certainly not formed by gaseous transmutation products because of its low Li-6-content, cannot be explained. Similar to the samples that were irradiated at high temperatures, also the pebbles that were irradiated at low temperatures do not show an observable trend as a function of the lithium enrichment, which is unexpected given significant differences in the tritium release rates [3]. The optical micrographs of these samples, however, suggest the existence of such a trend of increasing porosity with increasing Li-6-enrichment, in opposition to the density measurements by helium pycnometry. Compared to the influence of the irradiation temperature, an effect of the Li-6-content on the microstructure analyzed by SEM is less apparent. The pre-existing amount of about 10 mol% Li2SiO3 at the beginning increases during the irradiation due to the Li-loss as a result from the tritium breeding reaction and should be more pronounced with higher Li-6-contents. With regard to the same irradiation temperature, the SE micrographs show a comparable microstructure for samples with natural Li-6-abundance and samples enriched in Li-6. However, a difference can be observed compared to the depleted samples. Here, a distinctly lower amount of Li2SiO3 can be observed in the micrographs, which make sense in terms of a negligible tritium production and therefore an insignificant Li-loss. The effect of the Li-6-enrichment on the mechanical strength of the pebbles is in general also relatively mild. A slight decrease of the pebble strength with increasing Li-6-content is certainly visible. However, the irradiation temperature has a much stronger effect on the mechanical properties. This finding is also consistent with the presented observations in general. The mechanical strength is only expected to deteriorate, if significantly large defects are generated during the irradiation. These defects are seemingly generated by the inability of gaseous species to escape the pebble. As the release of tritium is apparently only mildly changed by the Li-6enrichment at low temperatures and virtually unchanged at high temperatures [3], the mechanical properties are changed accordingly. The nevertheless observable slight trend to lower mechanical strength with increasing Li-6-enrichment may be due to the increased formation and consequential segregation of Li2SiO3 at the grain boundaries of the Li4SiO4 grains which might facilitate crack propagation.

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the mechanical performance of non-irradiated but long-term annealed pebbles are compared to HICU samples. Apparently, the long-term annealing, which in this case was only performed for 96 days compared to about 400 days of irradiation within HICU led to the formation of different defects than the irradiation. This can be clearly seen by the low Weibull modulus that the long-term annealed sample shows compared to relatively high Weibull moduli of irradiated samples. Apparently, the long-term annealing leads to a significant growth of surface defects, which is very likely correlated with significant grain growth [19,20]. For the irradiated samples in contrast, the strength determining defects seem to be located within the volume. Potentially the irradiation-induced defects, for instance the formation of pores, as well as the generation of additional Li2SiO3, hamper the growth of the grains considerably, so that the mechanical properties do not degrade as severely as during the long-term annealing. 4.3.2. Constraint of pebble bed The degree of constraint of the irradiated pebble beds seems to have no significant influence on the material’s behavior, as no clear effects could be detected within HICU PIE. However, in some highly constrained samples, some sort of flattening of the pebbles can be observed in optical micrographs that could result from a dissipation of the pre-compaction stress of 0.7 MPa by creep. 4.3.3. Synthesis route and raw materials Some samples were produced using Li2CO3 as a starting material to adjust an enrichment of 20 % or a depletion of 0.06 % of Li-6 compared to the samples with natural Li-6-content. CO2 gas release from remaining carbonate during the irradiation could lead to a higher porosity of the pebbles and therefore can affect the release behavior of the material. Samples enriched with 20 % Li-6 showed relatively high values for the closed and total porosity before as well as after the irradiation. However, the depleted samples show the lowest closed porosities before and after the irradiation compared to the other samples. As total porosities are comparable for all samples, the open porosities for the depleted samples are relatively high, because of the before mentioned dependency on closed porosity (cf. section 3.3). A much stronger impact on the porosity that is mainly ruled by helium and tritium gas production during the irradiation has the presumed carbonate layer that was probably formed due to a certain storage time in air. This Li2CO3 layer was formed in all samples independent from the raw materials. As the thickness of the layer varies for each pebble in each sample, it is not possible to estimate relative amounts of lithium carbonate and find a direct correlation to the porosities. With regard to the mechanical properties, sample (HT, HC, 7.5, ex LOH) showed the highest strength after irradiation. In addition to that, the strength of the sample did not degrade in the same way as that of the other HICU samples, given the fact that the respective as-fabricated pebbles showed no exceptionally high strength. Why these pebbles excel in terms of mechanical strength is not clear, as all other properties show no exceptional deviation from the values that are obtained for comparable samples. 5. Conclusions

4.3. Influence of other parameters 4.3.1. Irradiation dose Most of the samples were exposed to doses of 10e12 dpa (calculated damage in steel). As all but two samples were exposed to doses in this quite narrow range, no effect in material’s properties with respect to microstructure, porosity or mechanical stability could be observed. An interesting observation can be made when

The post-irradiation examination of the Li4SiO4 based samples of the HICU experiment was performed and the results are presented and discussed. In summary, it can be said that the irradiation temperature most significantly affects the development of the material properties. A high irradiation temperature seems to be favorable for the mechanical properties of the pebbles. Beside higher mechanical strengths, lower formed close porosities are

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observed in pebbles irradiated at high temperatures. It seems therefore sensible to choose a high mean operating temperature for Li4SiO4 based pebble beds in a fusion blanket. Both observations appear to be highly correlated to the quicker release of gaseous species during the irradiation at high temperatures [3]. With regard to the Li-6 content, the effects are very small in contrast to the effects of the irradiation temperature. It can be concluded, that with an increase of the Li-6 content up to 20 % the material properties are not deteriorated. Yet it is very questionable if this finding can be extrapolated to the 60 % enrichment usually anticipated for a DEMO reactor [21]. For the HICU pebble beds that were pre-compressed, no clear negative effect on the mechanical strength of the pebbles was observed. The major effect that could be correlated with the precompression was the observation of slight flattening at contact points between individual pebbles, which is most likely caused by slight creep. Yet, for all HICU samples, a considerable amount of material obviously fractured. Because the material had to be retrieved from the irradiation capsules by some force, it is not clear to which extend this fracturing is an effect of the irradiation, which irradiation conditions are causative for the fracturing and to which extend the pebbles fractured during the retrieval. However, it is assumed to be most likely that most fracturing was caused during the retrieval. Furthermore, the formation of a lithium carbonate layer on the pebble surfaces complicates the analysis of the pebbles significantly and very likely obscures delicate trends in the material behavior. The obtained results clearly suggest that the formation of the layer took place after the irradiation. The results of the porosity measurements are certainly noticeably affected by this layer. Yet, it is assumed that the major findings are robust enough to be valid. Declaration of competing interest None. CRediT authorship contribution statement M.H.H. Kolb: Conceptualization, Methodology, Validation, Formal analysis, Investigation, Writing - original draft, Visualization. J.M. Heuser: Conceptualization, Methodology, Validation, Formal analysis, Investigation, Writing - review & editing, Visualization. R. Rolli: Methodology, Investigation, Resources, Writing review & editing. H.-C. Schneider: Conceptualization, Resources, Writing - review & editing, Supervision, Project administration, Funding acquisition. R. Knitter: Conceptualization, Writing - review & editing, Supervision, Project administration, Funding acquisition. M. Zmitko: Data curation, Writing - review & editing, Supervision, Project administration. Acknowledgement The authors greatly acknowledge the experimental support of the operators of the Fusion Materials Laboratory (KIT). The work leading to this publication has been funded by Fusion for Energy under the contract F4E-FPA-380-A3-SG01. This publication reflects the views only of the authors, and Fusion for Energy

cannot be held responsible for any use which may be made of the information contained therein.

References [1] J.B.J. Hegeman, E.D.L. van Essen, M. Jong, J.G. van der Laan, J. Reimann, Thermomechanical behaviour of ceramic breeder pebble stacks for HICU, Fusion Eng. Des. 69 (2003) 425e429, https://doi.org/10.1016/S0920-3796(03)000863. [2] M. Zmitko, P. Vladimirov, R. Knitter, M. Kolb, O. Leys, J. Heuser, H.C. Schneider, R. Rolli, V. Chakin, S. Pupeschi, L. Magielsen, A. Fedorov, Y. Poitevin, Development and qualification of functional materials for the European HCPB TBM, Fusion Eng. Des. 136 (2018) 1376e1385, https://doi.org/ 10.1016/j.fusengdes.2018.05.014. [3] J.M. Heuser, M.H.H. Kolb, R. Rolli, H.-C. Schneider, R. Knitter, M. Zmitko, The HICU PIE results of EU ceramic breeder pebbles: tritium release properties, J. Nucl. Mater. (2020). https://doi.org/10.1016/j.jnucmat.2020.152024. In press. €ke, B. Speit, D. Sprenger, Production process of [4] W. Pannhorst, V. Geiler, G. Ra lithium orthosilicate pebbles, Proc.20th SOFT (1998) 1441e1444. [5] R. Knitter, G. Piazza, J. Reimann, P. Risthaus, L.V. Boccaccini, Fabrication and Characterization of Lithium Orthosilicate Pebbles Using LiOH as a New Raw Material, Proceedings of the CBBI-11, 2004, pp. 108e119. [6] C.M. Sciolla, S.C. van der Marck, HICU (333-01) - Assessment of Burn-Up, Dpa and Nuclear Heat: 20295/11.107312 LCI/CS/MK, 2011. [7] S. van Til, HICU: Final Report on Irradiation: EFDA Reference TW2-TTBB-004bD3a, 2011. [8] H. Zimmermann, Mechanische Eigenschaften von Lithiumsilikaten für Fusionsreaktor-Brutblankets, Kernforschungszentrum, Karlsruhe, 1989. €tzmann, Thermodynamics of ceramic breeder materials for fusion re[9] O. Go actors, J. Nucl. Mater. 167 (1989) 213e224, https://doi.org/10.1016/00223115(89)90444-3. [10] K.L. Johnson, Contact Mechanics, first ed., Cambridge University Press, Cambridge [Cambridgeshire], New York, 1987. [11] A.G. Schott, Optical Glass, Data Sheets, 2018. https://www.schott.com/d/ advanced_optics/ac85c64c-60a0-4113-a9df-23ee1be20428/1.11/schott-optical-glass-collection-datasheets-english-17012017.pdf. (Accessed 13 May 2019). [12] E.R. Dobrovinskaya, L.A. Lytvynov, V. Pishchik, Properties of sapphire, in: V. Pishchik, L.A. Lytvynov, E.R. Dobrovinskaya (Eds.), Sapphire, Springer US, Boston, MA, 2009, pp. 155e176. [13] P.H. Shipway, I.M. Hutchings, Fracture of brittle spheres under compression and impact loading. I. Elastic stress distributions, Philos. Mag. A 67 (1993) 1389e1404, https://doi.org/10.1080/01418619308225362. [14] D. Munz, T. Fett, Ceramics: Mechanical Properties, Failure Behaviour, Materials Selection, Springer Berlin Heidelberg, Berlin, Heidelberg, 1999. [15] R. Ross, Bias and standard deviation due to Weibull parameter estimation for small data sets, IEEE Trans. Dielectr. Electr. Insul. 3 (1996) 28e42, https:// doi.org/10.1109/94.485512. [16] H. Hirose, Bias correction for the maximum likelihood estimates in the twoparameter Weibull distribution, IEEE Trans. Dielectr. Electr. Insul. 6 (1999) 66e68, https://doi.org/10.1109/94.752011. [17] R.B. Abernethy, in: R.B. Abernethy (Ed.), The New Weibull Handbook: Reliability & Statistical Analysis for Predicting Life, Safety, Risk, Support Costs, Failures, and Forecasting Warranty Claims, Substantiation and Accelerated Testing, Using Weibull, Log Normal, Crow-AMSAA, Probit, and Kaplan-Meier Models, fifth ed., North Palm Beach, Fla., 2008. [18] ReliaSoft Corporation, Life Data Analysis Reference Book, Available under a Creative Commons Attribution-NonCommercial-ShareAlike 4.0 International License, 2015. [19] G. Piazza, J. Reimann, E. Günther, R. Knitter, N. Roux, J.D. Lulewicz, Characterisation of ceramic breeder materials for the helium cooled pebble bed blanket, J. Nucl. Mater. 307e311 (2002) 811e816, https://doi.org/10.1016/ S0022-3115(02)00983-2. [20] G. Piazza, J. Reimann, E. Günther, R. Knitter, N. Roux, J.D. Lulewicz, Behaviour of ceramic breeder materials in long time annealing experiments, Fusion Eng. Des. 58e59 (2001) 653e659, https://doi.org/10.1016/S0920-3796(01)005178. ndez, P. Pereslavtsev, First principles review of options for tritium [21] F.A. Herna breeder and neutron multiplier materials for breeding blankets in fusion reactors, Fusion Eng. Des. 137 (2018) 243e256, https://doi.org/10.1016/ j.fusengdes.2018.09.014.