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Experimental thermo-mechanical testing apparatus for a DEMO relevant blanket design of a ceramic cladded pin
Fusion Engineering and Design 17 (1991) 161-164 North-Holland
161
Experimental thermo-mechanical testing apparatus for a DEMO relevant blanket design of a ceramic cladded pin G. D e l l ' O r c o a n d G . C . B e r t a c c i
ENEA-CRE Brasimone, CPI-40032 Camugnano (BO), Italy
The technological effort in supporting the development of a DEMO relevant Helium Cooled Solid Breeder Blanket (HCSBB) design, with an arrangement of the breeder inside the cladding tubes (BIT), foresees the qualification of some ceramic materials as tritium breeders. Some in-pile experiments are already carried out or are in progress, aiming at testing the materials as tritium release and selected mechanical properties. The facility allows to carry out some out-of-pile thermomechanical fatigue tests on a significant portion of one cladded ceramic pin. The breeder heat generation and the stresses are simulated by an internal electrical resistor. The radial thermal gradient is regulated by means of a proper air cooling flow with helium similitude. The internal hydraufics of the purge flow is fully reproduced, both chemically and thermally, and monitored in order to investigate the hot temperature corrosion effects on the sheath, its interaction with the ceramics and the pellet fractures.
I. Introduction The design of a Helium Cooled Solid Breeder Blanket (HCSBB) with lithium based ceramic pellets and a poloidal arrangement in tubular sheaths initially proposed by E N E A [1] is up to now under development at C E A / E N E A according to the European Test Blanket Programme. The C E A / E N E A ceramic breeder reference material is y-LiA102 because its better tensile strength and the thermal conductivity behavior at high temperatures than Li20, Li2ZrO 3 and Li4SiO4, with an acceptable Tritium Residence Time (TRT) of less than 24 h at 430 ° C [2]. At the moment, the fabrication process of such ceramics [3] is already mature for an industrial production (pilot plant) [4]. The performance evaluation of these products consists of a micro-structural analysis, physical, chemical and thermomechanical characterizations, both in and out of reactors. Some in-pile experiments, leaded by C E A and E C N laboratories, are checking their neutron performances also involving the tritium release. The reactor expected breeder stresses, due to high temperature, thick geometry, low ceramic thermal conductivity and high nuclear heat generation, are remarkable at the limit of the ceramic tensile failure strength [5]. Moreover, the long life expected in a reactor at high temperature, and consequently the creep deformation and the thermal fatigue damage need some more accurate investigations before the expensive nuclear tests, 0920-3796/91/$03.50
and for a rapid feedback on the design and the relative production process. The only possibility, up to now, is to perform such tests on the basis of the burst cycle and the life duration expected on a next step reactor like N E T [6]. The experimental facility here presented is utilized for checking the integrity, on the contrary, the size of the microfractures produced during the simulated thermal tests. The first test was started as the end of March 1991, with a batch of the E N E A reference y-LiAIO 2 ceramic pellets [4]. The same material with the same production process is also foreseen for another nuclear test ( E X O T I C VI) [4].
2. Test section At the present status of design [7] the single blanket module is composed of a poloidal tubular pressure vessel with a 7-pin or 19-pin bundle of cladded rods depending on its location in the segment box of the blanket. The single pin contains a stack of hollow ceramic pellets, whilst the bundle is equipped with a spacer system (grids or helicoidal wires wrapped around the cladding tubes). The bundle can be shrouded and insulated by a baffle in order to provide an annular gap for the inlet of the coolant. The beryllium, as neutron multiplier, is placed in form of squared blocks, outside the pressure
© 1991 - E l s e v i e r S c i e n c e P u b l i s h e r s B.V. A l l rights r e s e r v e d
162
G. Dell'Orco and G.C. Bertacci / Testing apparatus ior blanket de,~lgn
vessel in back row zones, or in form of a n n u l a r blocks in between the vessel and its baffle in front row zones. T h e helium, entering from the top of the module, flows downwards cooling the beryllium blocks. On the b o t t o m of the module, it reverses and flows upwards cooling the breeder bundle. The tritium, produced inside the lithium based ceramics, is swept by an ancillary helium purge flow. After being fed on the top of the module, it flows d o w n w a r d s inside half of the rod b u n d l e and then it reverses flowing upwards up to the top before its outlet. The helium main flow works at a pressure of 5 - 6 MPa, with inlet temperatures of 250 320 o C and an outlet t e m p e r a t u r e of 530 ° C. The helium purge flow is doped with H 2 or H20, in order to improve the tritium extraction [8] a n d works at a pressure close to the coolant. O n the basis of these parameters a proper test section was designed. The test section in Fig. 1 has a length of 1 m of cladded rod of the first row module. The hollow pellets are m a d e of E N E A sintered y-LiAIO 2 with 80% in theoretical density, 4.0 m m inside diameter, 9.2 m m outside diameter, 8.0 m m in height. They are inserted in an AISI 316L tube sheath of 10.0 m m in outside diameter a n d 0.30 m m in thickness. The breeder heat generation, inducing multi-axial thermal stresses, is simulated by an electrical wire resistor inside the hollow pellets. Its power is set up on the base of the equivalence of the tangential thermal tensile
IIR
In
stress expected on the external radius of the pellet. this case of simulated conductive heat transfer 4cross the ceramics, the thermal gradient is only a little higher than the gradient induced u n d e r reaction conditions. However, it is regulated by an a n n u l a r air cooling flow, fed at high t e m p e r a t u r e ( 5 0 0 ° C ) . The coolant is supplied at pressure, t e m p e r a t u r e a n d flow rate simulating the actual helium heat transfer tit the same reactor reference cladding sheath t e m p e r a t u r e (600 o (,). The fatigue cycling is imposed by a microprocessor controller regulating the electrical supply to the resistor, at the same estimated worst reactor b u r n pulsed period (200 s pulse d u r a t i o n a n d 70 s off-burn) for 5000 cycles, following the N E T technological phase specifications [6]. The resistor is made of a K A N T H A L A1 wire of 2.8 m m in d i a m e t e r brazed at two copper cold ends. These are connected to the relatively helium tight electric feedthroughs on the top and the b o t t o m of the two flanged caps of the test section. The wire resistor is electrically insulated from the test section by using some sleatite ceramic pellets or similar manifolds. The pin is welded to a bellow which, by a proper screw nut traction system, allows the thermal expansion comp e n s a t i o n without other stresses. The whole pin assembly is flanged to an AIS1 304 pressure vessel 3 / 4 " in nominal d i a m e t e r providing an a n n u l a r gap for the heat exchange with the air cooling flow. T h e test section is
BE/LO~
[N
CLADDED
/
PIN
\ I~
.,,
, ,
E !§-1
S:REi ~_o! T,C~
i
!, 7ETA',,:,,' ~ \,~
PRESSgRE
,~£¢llON
A A
Fig. 1. Test section.
\~HEAT~
VESSEL
~. : l ! p : ,
163
G. Dell'Orco and G.C. Bertacci / Testing apparatus for blanket design
instrumented with some thermocouples ANSI K 0.25 mm in diameter inserted in two pellets in opposite radial holes in order to control the thermal flow imposed on the ceramics. The helium purge gas flows upwards, across the two caps, inside the pin under test before its outlet and discharge to environment. The hydraulics of the sweeping helium flow is controlled in order to verify such possible modifications due to the pellet fractures affecting its pressure drop and flow rate.
3. Test
The second loop supplies the sweeping helium gas, doped with 0.01% H20 from the bottles. The piping, in I tt open loop arrangement and made of AISI 304 z , works at a maximum pressure of 1 MPa, a maximum flow rate of 4 Nm3/h and a maximum test section inlet temperature of 420 o C. All the components, like valves, fittings, transducers, are helium leak tested. The loop instrumentations, for the set up and the control of the experiences, together with the plant alarms, are recorded by a suitable data acquisition system, based on a PC compatible hardware architecture.
facility
The experimental rig, Fig. 2, called HE-FUS2, consists of two different loops. The first loop is an open crossed air loop, made of AISI 304 1" ½ tube, for the cooling of the test section. The loop organization allows to work with a rotative air compressor and with a small electric heater, by using an economizer for the enthalpy recovery before the air discharge. The economizer is a prototypical compact heat exchanger with a plate fin surface of 30 m2, a power of 50 kW and a theoretical efficiency of 70%. The electrical heater is made of four 6-rod cylindrical bundles with a total power of 22 kW. The loop design conditions are: pressure 1.2 MPa; temperature 530°C. The air working conditions are: maximum pressure 0.8 MPa, maximum flow rate 150 NmS/h; maximum test section inlet temperature 500 ° C.
~
4. Test
procedure
and controls
The tests are organized in order to collect the results by testing hundred pellets each time. Before the tests a sample of the batch of the ceramics is controlled by: - - geometrical and weight measurements; - - SEM analyses; - - compressive strength measurements; - - X-ray controls; - - sound velocity measurements. On the other hand, the sheath chemical composition and its internal oxidation status are also controlled. The fatigue tests are to be carried out by the experimental rig HE-FUS2 for 5000 cycles and with control steps in 500 cycles intervals. These controls consist of
AIR COMPRESSOR
He-FItTER
IIm
He-F~W METER
WIRERESISTOR ~
AIR TANK iR
IITER II
TESTSECTION"---~~ mum
ZER 4FLOW # NETE I
C- -
CERAMIC~
~i He-HEATER
~Iu
~
AIR HEATER Fig. 2. H E - F U S 2
test facility.
?
He BOTTZES He-SNIFFER
164
G. Dell'Orco and G.C. Bertaeci / Testing apparatus ,~or blanket design
the disassembly of the test section from the pressure vessel, the X-ray examinations of the whole pin assembly and, if nothing it is happened, the restart of the test. In case of evidence of pellet ruptures, the complete dismount of the tube and its pellet stack, for further control, is foreseen. This means an investigation on the tube-pellet interactions, a visual control on the single pellets, some further X-ray controls, a SEM analysis or ultra-sonic sample investigations. At the end of the whole test, other the previous controls, will be carried out a chemical analysis on the internal oxidation layer of the sheath.
5. Conclusion The experimental apparatus presented here allows to test a single cladded pin with a metallic sheath and a stack of lithium based ceramic hollow pellets. The outof-pile test consists of a simulated thermal fatigue stress, with reference to a fusion reactor expected pulsating cycle. The tests are i m p o r t a n t in order to validate easily and h a n d y the pin design and the fabrication process of the ceramics. The first results will be available in the middle of 1991. The future tests will regard other lithium based ceramics (LizZrO3) with martensitic steel sheath (MANET).
References [1] L. Anzidei, M. Gallina, L. Petrizzi, V. Rado, O. Simboiotti, V. Zampaglione, V. Violante and S. Bassani, Helium cooled
ceramic breeder in tube blanket for a tokamak reactor: the coaxial poloidal module concept design, Proceedings of Fusion Technology-15th SOFT, Utrecht, 1988 (Elsevier, Amsterdam, 1989) pp. 1194 1198. [2] J. Charpin, F. Botter, M. Briec, B. Rasneur. E. Roth and N~ Roux, Investigation of y lithium aluminate as tritium breeding material for a Fusion Reactor Blanket, Fusion Engineering and Design 8 (1989) 407 413. [3] C. Alvani, S. Casadio, L. Lorenzini, G. Brambilla, The ENEA selected fabrication process of y-LiAIO 2 ceramic pellets for tritium breeding experiments in Fusion Technology, High Tech Ceramics (Elsevier, Amsterdam, 1987). p p 2941 2948. [4] C. Alvani, L. Bruzzi, S. Casadio, L. Contursi, M.R. Mancini. A. Moauro, I. Stamenkovic, Fabrication and characterization of ¥-LiAIO 2 and Li2ZrO 3 ceramic breeder pellets for irradiation and blanket engineering experiments, presented at ISFNT-2, Karlsruhe, 1991. [5] J.K. Kiichle, Material data base for NET Test Blanket design studies, Blanket Advisory Group, November 1988. [6] The Net Team, Net Status Report, Eur-FU/XII-80/88-84. December 1987. [7] E. Proust, L. Giancarli, X. Raepsaet, J. Szczepanski, L Baraer, B. Bielak, F. Gervaise, J. Mercier, F. Vallette, L. Anzidei, P. Cecchi, S. Cevolani, M. Gallina, L. Petrizzi, V. Rado, V. Violante, V. Vettraino and V. Zampaglione, Status of the design and feasibility assessment of the European Helium-Cooled Ceramic Breeder Inside Tubes Test Blanket, Proceedings of 9th Topical Meeting on the Technology of Fusion Energy, Chicago, 1990. [8] C. Alvani, S. Casadio, V. Violante, M. Briec and M. Masson, In situ Tritium recovery from "y-LiAIO2 pellets: FEQUILA I, Proceedings of Fusion Technology 15th SOFT, Utrecht, 1988 (Elsevier, Amsterdam, 1991) pp. 1038 1045,
161
Experimental thermo-mechanical testing apparatus for a DEMO relevant blanket design of a ceramic cladded pin G. D e l l ' O r c o a n d G . C . B e r t a c c i
ENEA-CRE Brasimone, CPI-40032 Camugnano (BO), Italy
The technological effort in supporting the development of a DEMO relevant Helium Cooled Solid Breeder Blanket (HCSBB) design, with an arrangement of the breeder inside the cladding tubes (BIT), foresees the qualification of some ceramic materials as tritium breeders. Some in-pile experiments are already carried out or are in progress, aiming at testing the materials as tritium release and selected mechanical properties. The facility allows to carry out some out-of-pile thermomechanical fatigue tests on a significant portion of one cladded ceramic pin. The breeder heat generation and the stresses are simulated by an internal electrical resistor. The radial thermal gradient is regulated by means of a proper air cooling flow with helium similitude. The internal hydraufics of the purge flow is fully reproduced, both chemically and thermally, and monitored in order to investigate the hot temperature corrosion effects on the sheath, its interaction with the ceramics and the pellet fractures.
I. Introduction The design of a Helium Cooled Solid Breeder Blanket (HCSBB) with lithium based ceramic pellets and a poloidal arrangement in tubular sheaths initially proposed by E N E A [1] is up to now under development at C E A / E N E A according to the European Test Blanket Programme. The C E A / E N E A ceramic breeder reference material is y-LiA102 because its better tensile strength and the thermal conductivity behavior at high temperatures than Li20, Li2ZrO 3 and Li4SiO4, with an acceptable Tritium Residence Time (TRT) of less than 24 h at 430 ° C [2]. At the moment, the fabrication process of such ceramics [3] is already mature for an industrial production (pilot plant) [4]. The performance evaluation of these products consists of a micro-structural analysis, physical, chemical and thermomechanical characterizations, both in and out of reactors. Some in-pile experiments, leaded by C E A and E C N laboratories, are checking their neutron performances also involving the tritium release. The reactor expected breeder stresses, due to high temperature, thick geometry, low ceramic thermal conductivity and high nuclear heat generation, are remarkable at the limit of the ceramic tensile failure strength [5]. Moreover, the long life expected in a reactor at high temperature, and consequently the creep deformation and the thermal fatigue damage need some more accurate investigations before the expensive nuclear tests, 0920-3796/91/$03.50
and for a rapid feedback on the design and the relative production process. The only possibility, up to now, is to perform such tests on the basis of the burst cycle and the life duration expected on a next step reactor like N E T [6]. The experimental facility here presented is utilized for checking the integrity, on the contrary, the size of the microfractures produced during the simulated thermal tests. The first test was started as the end of March 1991, with a batch of the E N E A reference y-LiAIO 2 ceramic pellets [4]. The same material with the same production process is also foreseen for another nuclear test ( E X O T I C VI) [4].
2. Test section At the present status of design [7] the single blanket module is composed of a poloidal tubular pressure vessel with a 7-pin or 19-pin bundle of cladded rods depending on its location in the segment box of the blanket. The single pin contains a stack of hollow ceramic pellets, whilst the bundle is equipped with a spacer system (grids or helicoidal wires wrapped around the cladding tubes). The bundle can be shrouded and insulated by a baffle in order to provide an annular gap for the inlet of the coolant. The beryllium, as neutron multiplier, is placed in form of squared blocks, outside the pressure
© 1991 - E l s e v i e r S c i e n c e P u b l i s h e r s B.V. A l l rights r e s e r v e d
162
G. Dell'Orco and G.C. Bertacci / Testing apparatus ior blanket de,~lgn
vessel in back row zones, or in form of a n n u l a r blocks in between the vessel and its baffle in front row zones. T h e helium, entering from the top of the module, flows downwards cooling the beryllium blocks. On the b o t t o m of the module, it reverses and flows upwards cooling the breeder bundle. The tritium, produced inside the lithium based ceramics, is swept by an ancillary helium purge flow. After being fed on the top of the module, it flows d o w n w a r d s inside half of the rod b u n d l e and then it reverses flowing upwards up to the top before its outlet. The helium main flow works at a pressure of 5 - 6 MPa, with inlet temperatures of 250 320 o C and an outlet t e m p e r a t u r e of 530 ° C. The helium purge flow is doped with H 2 or H20, in order to improve the tritium extraction [8] a n d works at a pressure close to the coolant. O n the basis of these parameters a proper test section was designed. The test section in Fig. 1 has a length of 1 m of cladded rod of the first row module. The hollow pellets are m a d e of E N E A sintered y-LiAIO 2 with 80% in theoretical density, 4.0 m m inside diameter, 9.2 m m outside diameter, 8.0 m m in height. They are inserted in an AISI 316L tube sheath of 10.0 m m in outside diameter a n d 0.30 m m in thickness. The breeder heat generation, inducing multi-axial thermal stresses, is simulated by an electrical wire resistor inside the hollow pellets. Its power is set up on the base of the equivalence of the tangential thermal tensile
IIR
In
stress expected on the external radius of the pellet. this case of simulated conductive heat transfer 4cross the ceramics, the thermal gradient is only a little higher than the gradient induced u n d e r reaction conditions. However, it is regulated by an a n n u l a r air cooling flow, fed at high t e m p e r a t u r e ( 5 0 0 ° C ) . The coolant is supplied at pressure, t e m p e r a t u r e a n d flow rate simulating the actual helium heat transfer tit the same reactor reference cladding sheath t e m p e r a t u r e (600 o (,). The fatigue cycling is imposed by a microprocessor controller regulating the electrical supply to the resistor, at the same estimated worst reactor b u r n pulsed period (200 s pulse d u r a t i o n a n d 70 s off-burn) for 5000 cycles, following the N E T technological phase specifications [6]. The resistor is made of a K A N T H A L A1 wire of 2.8 m m in d i a m e t e r brazed at two copper cold ends. These are connected to the relatively helium tight electric feedthroughs on the top and the b o t t o m of the two flanged caps of the test section. The wire resistor is electrically insulated from the test section by using some sleatite ceramic pellets or similar manifolds. The pin is welded to a bellow which, by a proper screw nut traction system, allows the thermal expansion comp e n s a t i o n without other stresses. The whole pin assembly is flanged to an AIS1 304 pressure vessel 3 / 4 " in nominal d i a m e t e r providing an a n n u l a r gap for the heat exchange with the air cooling flow. T h e test section is
BE/LO~
[N
CLADDED
/
PIN
\ I~
.,,
, ,
E !§-1
S:REi ~_o! T,C~
i
!, 7ETA',,:,,' ~ \,~
PRESSgRE
,~£¢llON
A A
Fig. 1. Test section.
\~HEAT~
VESSEL
~. : l ! p : ,
163
G. Dell'Orco and G.C. Bertacci / Testing apparatus for blanket design
instrumented with some thermocouples ANSI K 0.25 mm in diameter inserted in two pellets in opposite radial holes in order to control the thermal flow imposed on the ceramics. The helium purge gas flows upwards, across the two caps, inside the pin under test before its outlet and discharge to environment. The hydraulics of the sweeping helium flow is controlled in order to verify such possible modifications due to the pellet fractures affecting its pressure drop and flow rate.
3. Test
The second loop supplies the sweeping helium gas, doped with 0.01% H20 from the bottles. The piping, in I tt open loop arrangement and made of AISI 304 z , works at a maximum pressure of 1 MPa, a maximum flow rate of 4 Nm3/h and a maximum test section inlet temperature of 420 o C. All the components, like valves, fittings, transducers, are helium leak tested. The loop instrumentations, for the set up and the control of the experiences, together with the plant alarms, are recorded by a suitable data acquisition system, based on a PC compatible hardware architecture.
facility
The experimental rig, Fig. 2, called HE-FUS2, consists of two different loops. The first loop is an open crossed air loop, made of AISI 304 1" ½ tube, for the cooling of the test section. The loop organization allows to work with a rotative air compressor and with a small electric heater, by using an economizer for the enthalpy recovery before the air discharge. The economizer is a prototypical compact heat exchanger with a plate fin surface of 30 m2, a power of 50 kW and a theoretical efficiency of 70%. The electrical heater is made of four 6-rod cylindrical bundles with a total power of 22 kW. The loop design conditions are: pressure 1.2 MPa; temperature 530°C. The air working conditions are: maximum pressure 0.8 MPa, maximum flow rate 150 NmS/h; maximum test section inlet temperature 500 ° C.
~
4. Test
procedure
and controls
The tests are organized in order to collect the results by testing hundred pellets each time. Before the tests a sample of the batch of the ceramics is controlled by: - - geometrical and weight measurements; - - SEM analyses; - - compressive strength measurements; - - X-ray controls; - - sound velocity measurements. On the other hand, the sheath chemical composition and its internal oxidation status are also controlled. The fatigue tests are to be carried out by the experimental rig HE-FUS2 for 5000 cycles and with control steps in 500 cycles intervals. These controls consist of
AIR COMPRESSOR
He-FItTER
IIm
He-F~W METER
WIRERESISTOR ~
AIR TANK iR
IITER II
TESTSECTION"---~~ mum
ZER 4FLOW # NETE I
C- -
CERAMIC~
~i He-HEATER
~Iu
~
AIR HEATER Fig. 2. H E - F U S 2
test facility.
?
He BOTTZES He-SNIFFER
164
G. Dell'Orco and G.C. Bertaeci / Testing apparatus ,~or blanket design
the disassembly of the test section from the pressure vessel, the X-ray examinations of the whole pin assembly and, if nothing it is happened, the restart of the test. In case of evidence of pellet ruptures, the complete dismount of the tube and its pellet stack, for further control, is foreseen. This means an investigation on the tube-pellet interactions, a visual control on the single pellets, some further X-ray controls, a SEM analysis or ultra-sonic sample investigations. At the end of the whole test, other the previous controls, will be carried out a chemical analysis on the internal oxidation layer of the sheath.
5. Conclusion The experimental apparatus presented here allows to test a single cladded pin with a metallic sheath and a stack of lithium based ceramic hollow pellets. The outof-pile test consists of a simulated thermal fatigue stress, with reference to a fusion reactor expected pulsating cycle. The tests are i m p o r t a n t in order to validate easily and h a n d y the pin design and the fabrication process of the ceramics. The first results will be available in the middle of 1991. The future tests will regard other lithium based ceramics (LizZrO3) with martensitic steel sheath (MANET).
References [1] L. Anzidei, M. Gallina, L. Petrizzi, V. Rado, O. Simboiotti, V. Zampaglione, V. Violante and S. Bassani, Helium cooled
ceramic breeder in tube blanket for a tokamak reactor: the coaxial poloidal module concept design, Proceedings of Fusion Technology-15th SOFT, Utrecht, 1988 (Elsevier, Amsterdam, 1989) pp. 1194 1198. [2] J. Charpin, F. Botter, M. Briec, B. Rasneur. E. Roth and N~ Roux, Investigation of y lithium aluminate as tritium breeding material for a Fusion Reactor Blanket, Fusion Engineering and Design 8 (1989) 407 413. [3] C. Alvani, S. Casadio, L. Lorenzini, G. Brambilla, The ENEA selected fabrication process of y-LiAIO 2 ceramic pellets for tritium breeding experiments in Fusion Technology, High Tech Ceramics (Elsevier, Amsterdam, 1987). p p 2941 2948. [4] C. Alvani, L. Bruzzi, S. Casadio, L. Contursi, M.R. Mancini. A. Moauro, I. Stamenkovic, Fabrication and characterization of ¥-LiAIO 2 and Li2ZrO 3 ceramic breeder pellets for irradiation and blanket engineering experiments, presented at ISFNT-2, Karlsruhe, 1991. [5] J.K. Kiichle, Material data base for NET Test Blanket design studies, Blanket Advisory Group, November 1988. [6] The Net Team, Net Status Report, Eur-FU/XII-80/88-84. December 1987. [7] E. Proust, L. Giancarli, X. Raepsaet, J. Szczepanski, L Baraer, B. Bielak, F. Gervaise, J. Mercier, F. Vallette, L. Anzidei, P. Cecchi, S. Cevolani, M. Gallina, L. Petrizzi, V. Rado, V. Violante, V. Vettraino and V. Zampaglione, Status of the design and feasibility assessment of the European Helium-Cooled Ceramic Breeder Inside Tubes Test Blanket, Proceedings of 9th Topical Meeting on the Technology of Fusion Energy, Chicago, 1990. [8] C. Alvani, S. Casadio, V. Violante, M. Briec and M. Masson, In situ Tritium recovery from "y-LiAIO2 pellets: FEQUILA I, Proceedings of Fusion Technology 15th SOFT, Utrecht, 1988 (Elsevier, Amsterdam, 1991) pp. 1038 1045,