tauro: a ceramic composite structural material self-cooled pb—17li breeder blanket concept

journalnf nuclear

materials ELSEVIER Journal of Nuclear Materials 233-237 (1996) 1257-1261

Tauro: a ceramic composite structural material self-cooled Pb-17Li breeder blanket concept

A.S. P6rez Rarrdrez a,~ A. Caso d L. Giancarli a,* N. Le Bars b, G. Chaumat c J.F. Salavy a, j. Szczepanski '~

a CEA Saclay, DRN/DMT/SERMA, F-91191 Gifsur-Yvette Cedex. France 2 b CEA Saclay, DTA/CE2M/LECMA, F-91191 Gif-sur-Yvette Cedex, France

c CEA Grenoble, CEREM/SGM, 17, rue des Martyres, F-38054 Grenoble Cedex 9, France d Soci£td Europ£enne de Propulsion (SEP) Le Haillan, B.P. 37, F-33165 Saint M[dard en Jalles, France

Abstract

The use of a low-activation (LA) ceramic composite (CC) as structural material appears essential to demonstrate the potential of fusion power reactors for being inherently or, at least, passively safe. Tauro is a self-cooled Pb-17Li breeder blanket with a S iC /S iC composite as structure. This study determines the required improvements for existing industrial LA composites (mainly S iC/S iC) in order to render them acceptable for blanket operating conditions. 3D S iC /S iC CC, recently launched on the market, is a promising candidate. A preliminary evaluation of a possible joining technique for S iC /S iC is also described.

1. Introduction

For future fusion reactors improved safety is one of the major criteria. Looking for a passively safe reactor, the Tauro blanket concept has been developed by CEA [1]. It is a self-cooled breeder blanket design using a CC as structural material and Pb-17Li as both coolant and breeder/mult ipl ier material. It is based on the power reactor specifications defined for the SEAFP study [2] with 3000 MW of fusion power, leading to an average neutron wall load of 2 MW m 2 and 0.5 MW m 2 peak FW surface heat load. The reactor has 16 toroidal field coils, 48 outboard and 32 inboard blanket segments. The Tauro blanket design minimises the energy potentially available in the vacuum vessel for breaking the confinement and releasing radioactive materials in case of accident. It is essentially a S iC /S iC box in which Pb-17Li flows in the poloidal direction providing the cooling of the first wall and the other box structures. This paper focuses on the

* Corresponding author. Tel.: +33-I-69082137; fax: +33-1- 69082381; e-mail: [email protected].

L EU Fellow, Ph.D. programme IFN, UPM (Spain).

critical design issues in order to determine the material properties and design characteristics needed to ensure reli- able and safe operation.

2. Concept and material development

2.1. Present outboard segment design

Compared to the previous version [1] the present design foresees to poloidally split the outboard segment into three modules ( = 3.5 m high) in order to ease fabrication and temperature control in the poloidal direction, Fig. 1. The coolant enters the top zone of the module through a single tube (20 cm diameter). The flow is divided afterwards in 5 subflows, one for each submodule. Then the Pb-17Li flows poloidally downward in the thin FW channel (1.25 cm) at 1 ms ~, turns into the second channel at the bottom, flowing up, down and up again (v = 0.06 ms i) towards the outlet collector, which combines the 5 submodule flows.

Calculations are performed for the mid-plane region where the thermal conditions are expected to be the most severe. The Pb-17Li headers have been defined in some detail in order to identify the needs in terms of joining.

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1258 A.S. P~;rez Ramirez et al./Journal of Nuclear Materials 233-237 (1996) 1257 1261

2.2. LACC candidate materials

The fabrication assessment has been perfonned for a 3D S i C / S i C composite recently launched on the market by SEP. which represents an important step towards the acceptability of the material. Nevertheless, the design calculations have been performed with the available properties for industrial 2D S iC /S iC already developed by SEP [I.3].

One important point is the behaviour of the S i C / S i C CC under neutron irradiation. Recent irradiations up to 25 dpa, in the temperature range 500°C to 1200°C [4], have shown a reduction in the mechanical and thernlat properties of the CC. Despite the initial reduction of the strength, the mechanical properties became independent of further neutron damage. The mechanism of the reduction in the thermal conductivity (3.) below 1200°C (phonon scattering at point defects, pores or small detect clusters), are related to those of tile swelling. The 3. is reduced with temperature and irradiation, but the effect of irradiation is recovered at higher temperatures. At low temperature, the irradiation reduction factor is roughly 4, while at 1200°C it is 1.2 [5].

A composite material made of SiC Nicalon fibers in SiO 2 matrix is studied by CEA as an alternative low activation solution to S iC /S iC [6]. Thick, dense S i C / S i O 2 samples can be obtained by an infiltration technique fol- lowed by hot isostatic pressing (HIP) for densification. At this stage, the process has to be optimised in order to decrease SiO 2 content. Uniaxial hot pressing technique will soon be used for densification instead of HIP, which may damage the long Nicalon fiber. Optimised samples

will be characterised for mechanical properties up to 1000°C.

2.3. Preliminary fabrication assessment

A preliminary fabrication assessment of Tauro has been performed by industry. With regard to materials, main difficulties are to be found in strongly curved parts, stiffeners and in the front double wall. In the present state of knowledge, it seems impossible to manufacture these complicated parts with usual industrial 2D-texture-based S iC /S iC CC, because of shaping difficulties and delami- nation problems.

As a consequence, the use of the structural 3D S iC /S iC material, is proposed. This new material, because of its multilayer texture, can be easily shaped without any de- lamination problem. This improved 3D material is based on the classic NL207 Nicalon fiber. Thicker, if needed, than a 2D material, it has also an improved surface finish because of the grain of its unique GUIPEX '~; texture.

Mechanical characteristics of the 2D and the 3D materials at room temperature compare as follows (2D/3D) : tensile stress 2 8 5 / 3 0 0 MPa: tensile strain 0 .75/0 .80%; Young's modulus 2 0 0 / 2 0 0 GPa.

The maximum thickness of the 3D material is approximately 6 mm. In the present state-of-the-art, it appears indeed very difficult to weave and densify thicker materials. For the Tauro application, a compromise has to be found between specifications (from 10 to 15 mm thickness) and the technical capabilities. Nevertheless, we do

Fig. 1. Present design of the Tauro blanket concept, outboard module.

A.S. P~rez Ramirez et al./Journal of Nuclear Materials 233-237 (1996) 1257-1261 1259

hope that, in medium term, it will become possible to manufacture thicker materials.

Deformability tests have been successfully performed and have proven the interest of the new 3D texture in achieving very complicated shapes such as U and comers. Considering these experiments, the curved front of the Tauro blanket appears to be feasible. Textile sewing and st icking/coinfi l trat ion are technologies that are yet to be studied in order to connect the stiffeners to the external

stmcture. With regard to collectors, their diameter ( = 180 ram)

makes them achievable with classic braiding technologies. The pipe connection design is yet to be studied. Owing to manufacturing reasons, a carbon composite will be prefer- able to S iC /S iC for these specific parts.

2.4. Joining of S iC/S iC CC

In order to use S i C / S i C as the structural material for the Tauro concept a suitable joining technique is needed. In particular, it is necessary to set up C C - C C joints capable to bear up to 1100°C (the upper limit of temperature to avoid CC degradation).

CEA has gained a 5 year experience in the field of refractory brazing of bulk SiC. From this background it is proposed to adapt the brazing process to the S i C / S i C case. Original filler metals, named 'BraSiC' , based on a Si-base intermetallic composition, suitable for a SiC substrate have been developed [7]. The advantages of these filler alloys, compared to other industrial fillers, are the following: high temperature resistance; low activation materials, mainly constituted by Si; no chemical attack to SiC substrate during the brazing process; very good adherence to the interface (bending strength higher than 350 MPa at 1200°C); good wettability on SiC: the observed angle in a sessile drop experiment is 30 ° (90 ° is the upper limit angle to qualify a filler alloy).

The main difficulty to extend their applicability to CC comes from the significant and intrinsic porosity of these materials ( = 10% in S i C / S i C SEP grade). The good wettability of BraSiC on SiC-based substrates leads to a partial or total infiltration into the CC. Current investiga- tions focus on optimisation of the brazing cycle operating parameters to control the braze viscosity.

2.5. Tritium confinement

The SiC ceramic material has tritium pemleation properties that makes it a very good candidate as permeation barrier. Tritium solubility and diffusivity in SiC are very low [8], and are only increased by irradiation defects. For the Tauro blanket concept the tritium permeation does not represent an issue, nevertheless the permeation behaviour of the S iC /S iC CC and dependence on the porosity is not known yet.

3. Outboard segment analysis

This section recalls the main results of the performed analyses for the Tauro outboard segment. The details are given in [11].

3.1. Mechanical calculations

The assumed maximum allowable stress is 140 MPa for the primary stresses and 190 MPa for the secondary ones [1]. Mechanical stresses are mainly due to the Pb-17Li hydrostatic pressure (1.5 MPa). The calculation has been done with the finite element (FE) code CASTEM 2000 [9], using a 3D geometry, square meshes and an orthotropic monolayer model of the S iC /S iC CC. The required prop- erty values following the anisotropy axes are not always available. Although not fully satisfactory for CC, the yon Mises stresses are used to indicate equivalent stress levels.

Wall thicknesses are given in [6]. According to the CASTEM calculations the maximum yon Mises stress (45 MPa) appears in the first rectangular channel. This value remains well below the design limits, and the deformations are negligible. This fact shows that the mechanical properties of present S iC /S iC CC are suitable for Tauro type blanket, provided the adopted design criteria will be confirmed by more complete evaluations, including irradiation effects.

3.2. MIlD effects

One of the most attractive features of the Tauro blanket concept is that it uses an electrical insulator as structural material. This allows a high-velocity liquid metal flow in the strong magnetic field B. In the presence of this B, the velocity profile of the flowing liquid metal (LM) will be flattened, influencing the heat transfer coefficient, the temperature distribution and also the pressure drop in the blanket [ 10]. The absence of eddy diffusivity in the blanket leaves only conduction as the remaining mechanism for heat transport perpendicular to flow direction. Some experiments have revealed velocity fluctuations in insulating ducts. The effect of these fluctuations is an increase of the effective heat-transfer, but they have a limited effect due to a linear electromagnetic breaking [1 l].

The MHD pressure drop for fully developed flow in a rectangular insulator duct can be estimated [10] as A p = vB~ L o / M (where v is the coolant velocity, B 0 the magnetic field module, L the length of the channel, cr the electrical conductivity of the fluid and M the Hartmann number). For the Tauro FW channel this is approximately 0.15 MPa (assuming that its rounded shape can be approxi- mated by a rectangle). The 3D MHD pressure drops could be fitted to A P 3 D = 0.5M ~/2cr vB2a (a is the half width of the channel) [ 12]. For Tauro A P3D = 0.025 MPa, leading to a total MHD pressure drop, after summing up all turns, manifolds or bends, of --- 0.35 MPa.

1260 A.S. P~re: Ramirez et al./Journal of Nuclear Materials 233-237 (1996) 1257-1261

3.3. Thermal-hydraulics and thermomechanics

An extrapolation from ID neutronic calculations indi- cates that the tritium breeding self-sufficiency is ensured with 0.7 ill depth in the outboard module and 0.5 m in the inboard one. The corresponding radial heat deposition distribution has been used as input for the thermal analysis [6].

The P b - l T L i has in l e t /ou t l e t temperatures of 275°C/625°C. These temperatures ensure enough margin against the maximum allowable operating temperature of S iC /S iC (1100°C), and at the same time a high themml efficiency can be expected. The temperature field in the blanket segment has been computed with the 2D FE code DELFINE [9] taking into account the MHD effects on temperature distribution. The contact between the S i C / S i C and tbe P b - 17Li considers the effect of the Hartmann side wall layer. Conduction remains the prevailing heat transfer mechanism.

The I cm thick FW is the most loaded part of the structure, with both the highest surface heat flux (0.5 MW 2) and volume heating (11 MW 3). The maximum temperature is located in the bottom part of the module, and is very sensitive to the S i C / S i C A value. Assuming an homogeneous value of 15 W m ~ K ~ [l], the temperature reached at the bottom part of a central module is 922°C (with an average of 670°C); at this location for the Pb-17Li the maximum temperature is 490°C and the average is 445°C. Using a slightly higher value for A (20

W m t K ~) , the maximum temperature in the S iC /S iC FW is 835°C (632°C average). The maximum temperature reached in the FW is well below the operating temperature limit (1100°C), providing some margin against possible power excursions without damage to the structural material. In the back part of the outboard module structure, the temperature levels reached (maximum of 650°C and average of 625°C) do not present any problem.

The thermomechanical stresses have been estimated with two methods. The first one estimates the stress due to the AT/ in the wall, using the simple isotropic formula o-= E a A T / 2 ( 1 - u) leading to the value of 185 MPa for A = 15 W m ~ K ~, just in the allowed limit and 150 MPa for A = 2 0 W m ~ K ~. The second one estimates the stresses created in the structure by the 3D temperature field. This field is extrapolated from the 2D DELFINE calculation and introduced in an orthotropic model in CASTEM [9]. The resulting thermal stresses peak at the location of the maximum FW temperature with a value of 120 MPa, far below the allowed limit. Accounting for the combined effect of thermal plus pressure loads, the resulting von Mises maximum stress is 130 MPa, again within the limits. The obtained module deformation is 5.3 mm in the poloidal direction. With the assumed thicknesses the temperature gradient through the S iC /S iC FW are just about acceptable.

From these studies it is confirmed that the principal need of improving the A of LACC lies in the material development program for fusion.

925

875

825

775

~. 725

~ 675

625

575

525

475

\ \ \

\ \

\

\\1

! - - - - - S i C / S i C F W _ _ - - Ph-171 i I: R h ~ n n o l

t I S i C / S i C B a c k p a r t ,

e - - - ~ P b - 1 7 L i B a c k p a r t

ls 10s lmin 15rain lhour lday COOLING TIME AFTER LOFA

Fig. 2. Temperature distribution after LOFA.

A.S. P[rez Ramlrez et al,/Journal of Nuclear Materials 233-237 (1996) 1257-1261 1261

3.4. LOFA analysis

The loss-of-flow-accident (LOFA) is one of the envis- ageable accidents for every blanket. In case of a self-cooled blanket, it can be considered as one of the most severe accidents. The good low activation properties of the S iC /S iC (low afterheat) compared to those of metals allows us to expect easy accident handling. The require- ment is that a LOFA should not prevent restarting the reactor operation after repair.

We have used the scenario assumed for DEMO [13]: complete stop of the coolant at the accident initiation; full power deposition for 1 s after the accident initiation; after the first second, the volume heat source drops to only the afterheat produced by the activated materials and the surface heat flux decreases linearly to 0 in 20 s.

2D thermal transient calculations have been performed, using the above mentioned assumptions, in two blanket module locations: in the mid-plane (maximum heat sources) and in the bottom-plane (maximum initial wall temperatures). The afterheat calculations have be performed with the ACAB [14] and FISPACT codes [15].

The DELFINE calculations have been performed using the following boundary conditions: zero heat flow through all the external walls, except through the back plate, where a radiation to the shielding is assumed. The emissivity of the S iC /S iC CC is 0.8 [3], while for the shielding steel the used value is 0.3, usually adopted for non-treated steel. Between the material boundaries into the module, perfect thermal contact is assumed.

The results are plotted in Fig. 2. It can be seen that the maximum temperature is reached after the first 10 seconds ( - 950°C in the FW), while the equilibrium temperature is reached at approximately 675°C 15 rain after the accident, following the cooling of the whole structure by radiation to the shield. The main conclusion is that the maximum temperature remains below the design limit, which means that, within this scenario, a LOFA can be handled without significant safety consequences. The maximum temperature strongly depends on the initial temperature and this one is well below the maximum limit.

4. Conclusions

The performed studies illustrate that Tauro blanket is a promising proposal for a passively safe blanket. The use of Pb-17Li as a coolant considerably reduces the energy available for accidental radioactive material release.

The new 3D SEP material allows a design using com- plex-shaped components owing to its texture and properties together with the prospective of specific S iC /S iC joining development. Some complementary development on evolved SiC weaving and sewing have to be performed.

The next step is to improve the high temperature mechanical properties of S iC /S iC composites with new high performance fibers such as High-Nicalon of Nippon Carbon.

One of the most important point for future R & D is the development of suitable joining techniques. Prospective studies have tested the applicability of a method developed for the SiC bulk material joining. These studies show that the joining of S iC /S iC CC by brazing can be envisaged once braze viscosity control can be achieved.

The thermal conductivity of the S i C / S i C structural material requires at least a 50% improvement of its present value. Another important issue is the porosity influence on the tritium tightness. The behaviour of S i C / S i C composite under irradiation needs be established.

References

[I] A.S. Prrez et al., Proc. of the 18th Syrup. on Fusion Technol., SOFT, Karlsruhe, Germany, August 22-26, 1994 (North Holland, 1995)pp. 1317.

[2] I. Cook, Fus. Eng. Des. 25 (1994) 179. [3] P. Fenici and H.W. Scholz, J. Nucl. Mater. 212-215 (1994)

60. [4] G.W. Hollenberg et al., J. Nucl. Mater. 219 (1995) 70. [5] S. Sharafat et al,, Fus. Eng. Des. 29 (1995) 411. [6] A.S. Prrez et al., Low-Activation Ceramic Composites as

Structural Material for Fusion Reactor Blankets, CEA/DMT Rept. (1994).

[7] A. Gasse, G. Chaumat et al., Brazing of SiC using Cu Si Non-Reactive Alloys, Proc. of the 4th Int. Conf. on Brazing, High Temp. Brazing and Diffusion Bonding, Aachen, Ger- many, 27-29 June, 1995.

[8] R.A. Causey et al., J. Nucl. Mater. 203 (1993) 196. [9] CASTEM Codes System, DRN/DMT, CEA Saclay, 1994.

[10] X.R. Wang et al., Fus. Eng. Des. 24 (1994) 389. [11] A.S. Prrez et al., Progress on the design of the Tauro

breeding blanket concept, CEA/DMT Rept. (1995). [12] T.Q. Hua and Y. Gohar, Fus. Eng. & Des. 27 (1995) 696. [13] E. Proust et al., Fus. Eng. Des. 22 (1993) 19. [14] J. Sanz et al., ACAB: activation code for fusion applications.

User's manual V 2.0, UCRL-MA-122002, August 1995. [15] R.A. Forrest and J.-Ch. Sublet, FISPACT4 User manual,

EASY Documentation Series UKAEA/FUS 287, AEA Technology, April 1995.

tauro: a ceramic composite structural material self-cooled pb—17li breeder blanket concept

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