Journal of Nuclear Materials 141-143 (1986) 311-315 North-Holland, Amsterdam 311

T E S T I N G N E E D S A N D E X P E R I M E N T S F O R S O L I D B R E E D E R B L A N K E T S *

P. G I E R S Z E W S K I t and R. P U I G H 2

Mechanical, Aerospace and Nuclear Engineering Department, University of California, Los Angeles, CA 90024, USA t On attachment from Canadian Fusion Fuels Technology Project ' Hanford Engineering Deoelopment Laboratory

Most of the critical issues for solid breeder blankets are related to the tritium and thermomechanical behavior of the solid breeder and multiplier. A major contributor to this uncertainty is the lack of definition of a preferred material, microstructure and form. Material selection and development is particularly cost-effective at present since much of the important blanket behavior is driven by local conditions rather than the overall blanket design, and since the choice of material affects the nature of the later and larger experiments. This stage of testing is also the most appropriate time to consider basic material choices such as sphere-pac or sintered pellet, and the incorporation of multiplier material into the blanket. The major testing needs and experiments are characterized here.

1. Introduction

The FINESSE study has recently characterized the major issues, experiments and strategies for the develop- ment of fusion nuclear technology [1]. The general classes of issues for solid breeder blankets are given in table 1.

The most important uncertainties are related to tri- tium breeding, tritium recovery, and breeder thermal behavior. These are particularly large for solid breeder blankets because: (1) there is limited understanding of gas transport in irradiated solids, (2) complex designs are used to keep the low thermal conductivity solids within their temperature limits under substantial nuclear heating and neutron damage rates, and (3) the resulting designs have a significant amount of non-breeding structure, coolant, and other material. The primary solid breeder safety uncertainties are the blanket tritium in- ventory, the tritium permeation rate into the coolant, and the breeder tritium and thermal behavior under transient conditions.

2. Testing needs

2.1. Tritium self-sufficiency

Most solid breeder blankets require 6Li enrichment and a neutron multiplier for adequate tritium breeding. Even so, within present uncertainties in data, modeling methods and design definition, it is not clear that pre- sent blanket concepts are self-sufficient in tritium [2].

The need for a neutron multiplier is a key issue for Li20. In all multiplied solid breeders, however, the tritium breeding is affected by the l~orm in which the multiplier is incorporated - which also affects the blanket tritium and thermal behavior. For example, the beryllium (n,2n) reaction produces low energy neutrons,

* Work supported by the U.S. Department of Energy, Office of Fusion Energy, contract DE-AM03-76SF00034.

so multiplication is enhanced if the breeder is mixed directly with the multiplier to minimize neutron absorp- tion in structure or coolant. However, this raises chem- ical compatibility concerns. An accurate assessment of the tritium breeding margin would thus indicate whether blankets without distinct multipliers were possible and, if not, what level of physical separation was acceptable.

2.2. Breeder~multiplier tritium inventory and recovery

Tritium transport and retention in the blanket are governed by a variety of phenomena. The major con- tributors to the total bJanket inventory are the tritium diffusivity, solubility and surface adsorption processes [3].

The uncertainty in the diffusivity can be much more than an order-of-magnitude, particularly at higher tem- peratures and burnups where there is almost no data. The influence on the tritium inventory is largest at lower temperatures since both diffusivity and the an- nealing of some radiation damage effects increase at higher temperatures.

The soluble tritium inventory is believed to be large only for Li20, where it is reasonably well measured [3]. In contrast, the surface inventory could be large for all breeder materials and is sensitive to surface conditions and the breeder chemical environment, particularly the oxygen activity. The oxygen activity can vary over many orders-of-magnitude depending on the controlling ther- modynamic system and the local reaction kinetics.

Table 1 General classes of solid breeder blanket issues

Tritium self-sufficiency Breeder/multiplier tritium inventory and transport Breeder/multiplier thermomechanical behavior Corrosion and mass transfer Structural response and failure modes in fusion environment Tritium permeation and processing from blanket

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312 P. Gierszewski, R. Puigh /Solid breeder blankets

Sufficient tritium is produced in the beryllium multi- prier to also be of concern (about 2 g / d in a 5000 MWth reactor with !2 cm Be [3]). The same tritium transport phenomena apply as with solid breeder materials, but there is very little data to assess their magnitude. And if the tritium must be removed by the coolant or purge streams, then there are either coolant contamination or breeder/multiplier chemical interac- tion concerns.

2.3. Breeder / rnultipfier thermomechanical behaoior

The thermal behavior of the breeder, and conse- quently the overall blanket design, is constrained by the relatively low thermal conductivity and upper tempera- ture limits presently assumed. Large variations in ther- mal conductivity are possible over the range of plausible breeder conditions [4]. The upper temperature limits depend on the effects of many processes such as sinter- ing, creep, phase change, vapor phase transport and corrosion. The uncertainties and experimental complex- ity of high temperature tests has led to a lack of experiments that could define the temperature limits.

There are also no completed tests that indicate the extent and consequences of breeder/cladding mechani- cal interactions or temperature gradients withiia the breeder or multiplier; even the breeder and multiplier mechanical properties are not well known. For example, even though beryllium has been used in fission reactors, the available high fluence swelling data is based on a few post-irradiation annealed specimens [5].

2.4. Corrosion and mass transfer

Chemical compatibility and mass transfer influence the temperature limits of the solid breeder and multi- plier. Most breeder ceramics are not particularly reac- tive with proposed cladding materials, except for Li 20 where vapor phase transport and corrosion are signifi- cant concerns. Other than ideal equilibrium thermody- namic calculations, there are no experiments to indicate the kinetics of in teract ions between possible breeder/multiplier mixtures, or the effects of high end- of-rife burnup on the local chemistry.

3. Experiments

The resolution of these and the other issues will require a range of experiments, from property measure- ments to integrated tests of full blanket modules. Fig. 1 summarizes the major classes of experiments for solid breeder blankets, and the solid breeder/multiprier ex- periments are briefly discussed below. More details may be found in ref. [1].

By appropriate 6Li enrichment, it is possible to simulate fusion tritium generation and heating rates in fission reactors. Most radiation-related effects can also be simulated - specifically burnup-related stoichiome- try changes, helium production, and damage due to the energetic tritium and helium fission products. Reactors with vented test capabilities can also provide direct simulation of the purge environment.

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Fig. 1. Types of experiments and facilities for solid breeder blankets. (Some experiments and facilities exist.)

P. Gierszewski, R. Puigh /Solid breeder blankets 313

3.1. Material development and characterization

Material development refers to the process of identi- fying possible materials, understanding the effects of material parameters on the properties, fabricating materials with the desired material parameters, and characterizing the material through measurement of its properties. Material parameters include the crystal form, grain size, pore size, physical form, impurity or additive content, phase purity, and fabrication process.

The most important needs at present are for basic properties for all compounds (particularly tritium diffu- sion, tritium adsorption, thermal conductivity, swelling, and thermal stability), and the fabrication of controlled microstructure specimens (including sphere-pac).

3.2. Tritium recovery experiments

The most relevant tests involve irradiation to provide internal tritium generation, heating and fluence effects. These can be closed or open capsule tests using isother- mal specimens, pellets large enough to support reactor-relevant temperature gradients (or to achieve high center temperatures), and /or pellets with signifi-

cant mechanical interaction with the container walls. The importance of an actively-controlled flowing gas environment has been demonstrated in recdnt experi- ments. However, closed capsule experiments are less expensive and provide useful scoping data. A number of open capsule irradiations (table 2) are underway or have been completed. These tests are exploring a range of temperatures, temperature gradients, breeder materials, container materials, burnups and sweep gas composi- tions and flow rates.

However, the planned tests will not address the combination of moderate-to-high burnup with a flowing purge gas under temperature gradients and breeder/cladding interactions. Although these effects will be considered separately, synergistic effects and modelling inaccuracies will make extrapolation to reac- tor-relevant combinations uncertain. Consequently, the next major class of tests should address these interac- tions. Such advanced in-situ tritium recovery experi- ments could still be performed with in-core capsules. The importance of achieving significant burnup while limiting self-shielding in a fission reactor neutron spec- trum leads to relatively long irradiation times and a preference for fast reactors.

Table 2 Completed and active in-situ tritium recovery irradiation tests

Experiment Ceramic Grain Density Temperature Lithium Container Time size (7oTD) (K) burnup frame (~m) (atTo)

TRIO LiAIO 2 0.2 65 400-700 0.2 SS 84/85 (US) (9 mm thick annular pellets)

VOM-15H Li 20 10 86 480-760 0.24 - 84 (Japan)

VOM-22/23 Li 2 ° - - 400-900 0.04 SS 86 (Japan) (4 mm diameter pellets)

LiAIO 2 - - 400-900 0.1 SS 86 (4 mm diameter pellets)

LILA LiA102 0.4-13 78 375-600 0.02 Quartz, 86 (France) (10 mm diameter pellet) SS

LISA Li 2 SiO3 30-80 87 450-730 - SS 86 (Germany) (10 mm diameter pellet)

Li 4 SiO4 26 94 450-730 - SS 86 910 mm diameter pellet)

LiA102 0.4 78 450-600 - SS 86

EXOTIC LiAlO 2 1,8 80,90 400,600 0.1 SS 86 (Neth./UK/ 914 mm diameter pellet) Belgium)

Li 2SIO3 1,12 80,95 400,600 0.1 SS 86 (14 mm diameter pellet)

Li 20 5-10 80 400,600 0.1 SS 86 (14 mm diameter pellet)

CRITIC Li 20 20 80 400-950 0.15 Inconel- 86 (Canada) (10 mm thick annular pellet) 600

314 P. Gierszewski, R. Puigh / Solid breeder blankets

3.3. Breeder thermomechanics experiments

Although unirradiated tests of mechanical properties can be performed relatively easily with standard equipment, the important breeder/cladding interactions and breeder thermomechanical behavior are affected by rad ia t ion (e.g., swelling, creep) and larger geometrical/operating effects (e.g., settling, cycling, cracking). The radiation effects can be determined in the same tests as those for tritium recovery, and scoping tests with temperature gradient and breeder/cladding interactions are underway (e.g., FUBR-1B). However, several closed capsule tests dedicated to thermomecha- nical effects are needed in order to allow complete instrumentation (e.g., thermocouples distributed inside the solid breeder).

mark design codes, and study severe transients. It is much more difficult and expensive to perform such tests in irradiation facilities.

Nuclear test assemblies for fission reactors can pro- vide the maxim concept verification possible in non- fusion devices. Such tests would include the important nuclear effects but be limited in other respects. A full blanket module would need about 1 m 3 test volume, and only achieve the equivalent of (at most) a i M W / m 2 heating rate in any existing reactor. In-core assemblies could be placed in existing fission reactors at reactor relevant heating rates (2-5 M W / m 2 neutron wall load), but would be limited to about 10 cm diameter.

3.4. Corrosion and mass transfer experiments

Experiments to determine temperature limits based on material interactions involve long-term tests at con- ditions which are achieved in many of the tritium recovery experiments. However, for new and /o r more reactive materials, separate unirradiated testing at reac- tor temperatures can provide cost-effective data to judge the feasibility of the material or to provide controlled test conditions for model development. Useful tests include mass transfer within and from Li 20 in a purge stream with' hydrogen, and the interaction kinetics of beryllium with solid breeder and clad.

3.5. Multiplier behavior experiments

For beryllium or other solid neutron multipliers, the mechanical behavior and tritium retention under reac- tor conditions are significant uncertainties. Experiments needed include unirradiated property measurements, and irradiated closed and open capsules as with the solid breeder. Fission reactors-such as FFTF can pro- vide reactor-relevant simulation of helium and tritium production in beryllium.

3.6. Partially integrated experiments

More complex tests with more relevant geometry, size and environmental conditions can provide some concept verification information. Non-neutron test stands, fission reactors and fusion devices can serve different roles. However, only a fusion device can pro- vide fully integrated testing.

Non-neutron thermomechanical tests with non- nuclear heat sources can be used to test up to full blanket modules. Although there are clearly limitations on the simulation of reactor heating.profiles and irradi- ation effects, these tests provide an opportunity to explore complex thermomechanical behaviors (e.g., gap conductance, flow distribution, thermal cycling), bench-

Table 3 Major solid breeder blanket tasks

Solid breeder material development and characterization - Measurement of tritium retention and release, including

effects of burnup, material characteristics and purge flow chemistry;

- Thermophysical and thermomechanical properties, including effects of irradiation and material characteristics;

- Development of sphere-pac material; - Assessment of novel materials, particularly those that com-

bine breeder and multiplier; - Development of fabrication and recycling techniques.

Multiplier material development and characterization - Measurement of swelling in beryllium irradiated at temper-

ature, including effects of form and porosity; - Measurement of tritium retention and release, particularly

the effects of form and irradiation; - Measurements of irradiation creep and mechanical proper-

ties; - Development of Iow-loss-rae fabrication and recycling tech-

niques.

Blanket thermal behavior - Measurements of corrosion, mass transfer and chemical

interaction kinetics, particularly for Li20 and beryllium- containing materials;

- Measurements of breeder/multiplier temperature profile and thermomechanical effects of breeder/cladding interaction;

- Non-neutron blanket (sub)module thermomechanical integr- ity, including cycling, corrosion, normal transients, and severe transients.

Neutronics and tritium breeding - Simple geometry mockups for important blanket material

combinations; - Engineering mockups of blanket designs and adjacent reac-

tor sector.

Advanced in-situ tritium recove~ - Two or more instrumented and purged assemblies with

multiple capsules.

Nuclear submodule experiments - Two or more nuclear submodule assemblies.

P. Gierszewski, R. Puigh / Solid breeder blankets 315

4. Test plan

Based on the key issues and testing needs, a number of broad tasks have been identified for solid breeder blankets [1]. These are summarized in table 3. Over the next 15 years, the task emphasis should shift from understanding of materials behavior, to developing pre- dictive capabilities and, finally, to verifying the design concepts. Accordingly, the development and characteri- zation of solid breeder materials must continue. In the near future, additional tasks must be started to develop and characterize a neutron multiplier material (prob- ably beryllium) in a form that supplies sufficient tritium breeding and is compatible with the solid breeder and /o r the tritium recovery system. The design of ad- vanced in-situ tritium recovery experiments should also begin, in order to quantify local design-related behavior under combined fusion-relevant conditions. From this data base, a limited number of blanket concepts would be selected and verified to the extent possible in non-fu- sion facilities. This test plan would support an assess- ment of the feasibility and attractiveness of solid breeder

blankets within the next 15 years at an estimated cost of 10-20 M$/yr [1]. Assuming that solid breeder blankets are sufficiently attractive, the program would then be able to confidently proceed with a solid breeder blanket experiment in a fusion test device.

References

[I] M.A. Abdou et al, Technical issues and requirements of experiments and facilities for fusion nuclear technology, FINESSE Phase I Report, University of California at Los Angeles, PPG-909/UCLA-ENG-85-39 (December 1985).

[2] M.A. Abdou et al, Deuterium-tritium fuel self-sufficiency in fusion reactors, Fusion technol. (1986).

[3] D.L. Smith et al, Blanket comparison and selection study (Final Report), Argonne National Laboratory, ANL/FPP- 84-1 (1984).

[4] Y.Y. Liu and S.W. Tam, Thermal conductivities for sintered and sphere-pac Li20 and LiAIO 2 solid breeders with and without irradiation effects, Fusion Technol. 7 (1985) 399.

[5] Beryllium Technology Session (6th Topical Meeting on the Technology of Fusion Energy, San Francisco, March 1985), Fusion Technol. 8(1) (1985) Part 2.