A blanket concept with liquid Li17Pb83 for tritium breeding in INTOR-NET

Nuclear Engineering and Design/Fusion 1 (1984) 185-194 North-Holland, Amsterdam

185

A B L A N K E T C O N C E P T W I T H L I Q U I D LitTPbs3 F O R T R I T I U M B R E E D I N G IN I N T O R - N E T J. A I R O L A , M. B I G G I O , G. C A S I N I , F. F A R F A L E T T I - C A S A L I , P. LI BASSI, C. P O N T I , M. RIEGER Commision of the European Communities, Joint Research Centre, lspra Establishment, 21020 Ispra (Va), Italy and C. P I A N A University of Milan, Milan, Italy

A blanket concept with eutectic LilTPbs3as liquid breeder, suited for tritium production in an experimental Tokamak power reactor is outlined and discussed. This design has been developed to satisfy the INTOR-Phase-I specifications,in particular: (i) modular arrangement of the blanket units inside the vacuum vessel; (ii) no use of the heat deposited for electricity production, (iii) a net tritium breeding of at least 60%. In this article the main results of the neutronics and thermohydraulics analysis are reviewed and the problems identified. Methods to keep liquid in the breeder during operation are proposed and discussed. The consequences of a coolant tube rupture in a breeder unit appears to be the most serious problem.

1. Introduction

The investigation of the eutectic LilvPbs3 as liquid breeder in experimental and commercial fusion power reactors started a few years ago. The main reason for the interest of the scientific community in this material was related to its expected low chemical reactivity with air and water, as compared to the case of lithium metal. Other features are: - inclusion of a neutron multiplier, in the breeder itself, which enables high tritium breeding ratios to be obtained; - low solubility of the hydrogen isotopes which lead to a low tritium inventory in the breeder. Possible disadvantages are connected with: - the high melting point (235 ° C), - the high density ( = 10 g/cm3). In general the knowledge of the basic properties of the eutectic LilTPbs3 was found to be poor. As a consequence an experimental investigation was started a few years ago at JRC-Ispra to implement a data base for this breeder material. Meanwhile, in the frame of the I N T O R / N E T studies, a conceptual design of a blanket using liquid LilTPbs3 as breeder has been produced. The possibility

of using this material as a breeder for demonstration and commercial power reactor is being pursued in other laboratories (CEA-Saclay [1], ANL [2], University of Wisconsin [3]). In the following an overview of the engineering aspects of the I N T O R / N E T breeding blanket developed at JRC-Ispra is presented. The proposed methods for tritium recovery from the blanket are discussed in another article of this issue.

2. D e s i g n specificaiions and blanket description

According to the specifications of the INTORPhase-I workshop [4], a number of assumptions have been made to guide the conceptual design study, namely - first wall separated from the breeder region, - water as a coolant both for first wall and breeder region, - no electricity production, - inclusion of a tritium breeding blanket only in the outer and top part of the reactor. The segmentation and assembling of the first wall/blanket units inside the reactor have been conceived in order to fit with the general criteria for the mechanical configuration and maintenance presented in

0 1 6 7 - 8 9 9 x / 8 4 / $ 0 3 . 0 0 © Elsevier Science Publishers B.V. ( N o r t h - H o l l a n d Physics Publishing Division)

186

J. Airola et al. / A blanket concept with liquid LijTPbss

another article of this Issue. The following design features have been selected: the first wail-breeding blanket system is located inside the main toroidai vacuum vessel of the machine; - it is subdivided into 24 segments which can separately be removed from the reactor through access ports for replacement and maintenance; - each segment includes 6 breeding units along the toroidal direction and 4 breeding units along the poloidal one. In total each blanket segment contains 24 breeding blanket units of similar size as indicated in fig. 1. Each reactor segment is completed by a shielding region in its inner and lower part and by the plug, which closes the segment inside the access port. For each segment the first wall is formed by a series of tubes facing the plasma and separated from the breeder region as shown in fig. 1. These tubes are arranged along the toroidal direction and cooled by heavy water. Fig. 1 presents the arrangement of the inlet/outlet collectors for the first wall cooling tubes as well as their attachment to the reactor segment. Note that the tubes are free for dilatation along the toroidai direction. The cross-section of each tube has a flat shape in front of the plasma in order to get a homogeneous erosion due to sputtering and to reduce the pumping requirement and the hydrogen desorption during start-up. The main parameters of the first wall are presented in table 1. The inlet/outlet temperatures of the coolant are low (50/100 o C, respectively) in order to limit the maximum temperature of the first wail to -

BREEDER COOLANT HIO OU BOARD BLANKET SEGMENTS WITH EUTECTIC U-Pb BREEDER I

=

._~~-J~-Itll

s.e

:!

SHELL FOR PASSIVE J~ I=LARMAVERTICALSTABILISATION

'~ ~ l ~

"~ _~._n OF THE PLASMA

R

IJ

~I~.//.~, =~

_

-

FI

4.8--. ~

WATERCOOLED SUPPORT,NGP

\

-

~

TE

FIRST WALL

~¢'1:1 (RADIATION PANEL) LOOP FOR PASSIVEVERTICAL ..,.~_TABlUSATION OF THE PLASMA

OUTBOARD FIRST WALL COOLANT (D20) (INLET AND OUTLET) q

1

t~

Fig. 1. INTOR/NET blanket arrangement.

Table 1 First wall heat transfer parameters Coolant Total heat flux absorbed in first wall (surface heat load + neutronic heat in structure) Diameter of coolant channels Inlet temperature of coolant Outlet temperature of coolant Coolant velocity Convectiveheat transfer coefficient Coolant pressure

D20

4.2 x 105W/m2 0.01 m S0°C 100°C 1.9 m/s 1.3 × 104 W/m2K 1 MPa

350°C. This will avoid some of the bulk radiation damage effects, such as swelling, and increase the first wail lifetime. The design of the breeder units and their cooling structures result from a compromise between various requirements. In particular: - the minimum temperature and pressure of the coolant is determined by the need to ensure that the LilTPbs3 is always liquid during the reactor operation, - the thickness of the unit container is determined to satisfy the safety requirements in the case of a rupture of a cooling tube, - to avoid corrosion problems with the stainless steel structures, the maximum temperature of the breeder must not exceed 400 ° C. On this basis the breeder unit design shown in fig. 2 has been selected. Inside each vessel unit a tubular heat exchange system is completely immersed in the liquid breeder. The penetrations for coolant inlet and outlet tubes are placed on the rear side of the unit. For each blanket segment all cooling tubes are supplied in parallel from one main coolant inlet duct. All the water outlets lead to a main outlet tube in the same way. The main supply and return cooling lines of each segment are situated behind the mechanical structures where all the breeder units are supported. The main parameter of the breeder units are given in table 2. In each blanket segment the blanket units are supported by segment plug structures through hinges attached to metallic plates. In the case of a plasma disruption the hinges are designed to support the torques induced by the eddy currents in the breeder unit structures. They are also designed to support the weight of the breeder unit. Such an arrangement of the supporting system enables a possible substitution by remote handling of one of the breeder units, once the blanket segment is removed from the reactor. The liquid breeder

J. Airola et aL / A blanket concept with liquid Li /zPbsj

187

Table 2 Breeding blanket heat transfer parameters

s.s.~~ AISI 316.

,,

~'"'

~ i,~

Coolant Total heat flux absorbed in the blanket Dimensions of the cooling tubes in the blanket, outer diameter/inner diameter Inlet temperature of coolant Outlet temperature of coolant Coolant pressure Coolant velocity Convective heat transfer coefficient

-p~L~'GOOLANT EDER i)l~ H,O

LIQUID BREEDER EUTECTIC Li-Pb TM

?k.I I I I II

FIRST WALL (RADIATION PANEL)

/--SHELLFOR ~ PASSIVE

H zO 1.2 × 106 W/m 2 21/18 mm 240 o C 260 o C 5 MPa 3.3 m/s 2.5 x 104 W/m 2 K

",S+AB,L,SAT,ON ?OF THE PLASMA / / "

flt "~ ~WATER / COOLED

~ ~ '"J //~ ~ " ~'--" " '

/SUPPORTING --PLATE

LIQUID BREEDING BLANKET UNIT

4OO

Fig. 2. INTOR/NET blanket unit.

is continuously circulated at a low flow rate for tritium recovery into an extraction column outside the vacuum vessel. For each reactor segment, the breeder vessels are interconnected by tubes. Each unit has the breeder inlet at its lowest point and an outlet at the highest point (fig. 3). These tubes are connected by horizontal collector ducts in the rear zone, outside the blanket area. By joining the outlet collector from a lower unit level to the

inlet collector of the following higher unit level, all vessels can be supplied uniformly. The draining of the system is possible in the same way, in the reverse direction by gravity only. The filling and draining of the liquid breeder is done in situ. Therefore, for assembly and disassembly of the reactor segment, only the empty breeder units have to be moved. The pressure drop due to magneto-hydrodynamic effects in the breeder loop is small, since only a small mass flow rate is pumped to the extraction column. In fig. 4 a layout of the plant circuits related to the breeder and water coolant of the breeder region is shown. Before start-up, the breeder is pumped into the units in a liquid phase. For that purpose it is preheated in a storage tank clearly above the melting point at about 300 o C. The total structure of the breeder units is also preheated to about 250°C before filling, in order to avoid freezing of the breeder during that process. The tubes where Li]7Pb83 flows are all equipped with electric trace heaters fixed in the

4OO

So+O, o o

o"0

o o .®-© o

o o r-:>-

Fig. 3. Cross-section of the breeder unit.

+

J. Airola et al. / A blanket concept with liquid Li tTPb83

188

y-

WATER

Z

PRESSURISER PRIMARYLOOP

PRESSURISER SECOND. LOOP

INTERMEDIATE HEAT EXCHANGER

PUMP

HEA~

__

PUMP

PUMP

Uw Pba3 DUMP TANK

Fig. 4. Lay-out of the plant circuits.

longitudinal direction on the outer walls. This preheating method has already proven its reliability in radioactive liquid metal systems. Indeed, the purification circuits of the primary sodium loops of the Ph6nix fast fission reactor are preheated in this way. The vessels of the breeder units are preheated by hot gas circulating in the space provided to contain the breeder. For that purpose, a bypass is provided in the circulating loop of the breeder which is equipped with a heating system for inert gas and a blower to circulate it across the unit vessel structures. The total cooling system inside the breeder units is also preheated before the filling takes place to its working temperature. For this purpose an electrical heater is installed in a bypass of the secondary loop of the water cooling system. This heater also provides the necessary energy to keep the breeder in the liquid state during the shut-down periods of the operating cycle. A preliminary analysis of the coolant accidents related to this type of blanket has been carried out by Klippel [5]. The main results are as follows: the first wall remains safe in case of loss-of-flow (LOFA) and loss-of-coolant accident (LOCA), the temperature response of the breeder units in case

of LOFA and LOCA is relatively slow so as to allow the intervention of the safety systems, - a pipe break (LOCA) inside the vessel units causes pressure oscillations with a peak of about 9 MPa and a static final pressure equal to the coolant pressure (6.5 MPa) after about 5 ms. These results have been obtained with one-dimensional calculations and they are being checked with more refined two-dimensional calculations and with the results of experiments in progress at JRC-Ispra. Measures to mitigate the fluid-structure interaction, e.g. rupture disks will be investigated.

3.

Neutronics

3.1. Survey calculations Neutronics calculations have been performed to determine the TBR (tritium breeding ratio), the heat deposition and the energy multiplication. A first set of survey calculations have been performed to understand the basic features and to look for an optimum set of design parameters, from the viewpoint of the TBR. In

J. Airola et al. / A blanket concept with liquid Li lTPbs3

particular these calculations have provided information on:

the optimum value of the water content in the breeding region; the effect of increasing the amount of the structural material; the effect of the first wall; - the need for lithium enrichment. The highest TBR is obtained when the volume fraction of water is between 10 and 12%. This quantity is somewhat higher than the minimum required for cooling. In these calculations the density of water has been assumed to be 1 g / c m 3 (room temperature). If the amount of stainless steel structural material present in the breeding zone is increased from 9 to 19% by volume, the value of the TBR decreases by 0.10. The sensitivity of the TBR to the amount of structural material is hence -0.01 for each 1% increase of stainless steel, and is constant in the interval considered. The effect of cooling the first wall with light or heavy water, and of varying the thickness of the structural material, is shown in table 3. These calculations have been performed assuming a homogeneous composition of the first wall, with a total thickness varying between 10 and 25 mm, in which the coolant always has an equivalent thickness of 4 mm. An increase of 1 mm in the thickness of stainless steel causes a decrease of about 0.01 in the TBR. Replacement of H20 with D20 leads to an increase of the TBR by 0.07. These data emphasize the dramatic effect of the presence of structural material and water in front of the breeder material. They must be as thin as possible if a satisfactory value of TBR is to be obtained. The effect of steel is twofold: it slows down the neutrons by inelastic scattering reactions, which reduces the multiplication by lead, and it captures the slow neutrons (in the eV range) which have been further slowed down by the water. The amount of water in the first wall should be the minimum needed for cooling requirements. Lithium burn-up due to neutron absorption by Li 6 is about 5% of the Li atoms initially present in the eutectic

-

-

-

189

Li17Pbs3. This is the main reason why the natural enrichment is not sufficient. The quantity of Li 6 in the breeding region would decrease sharply with a consequent decrease in the TBR. An enrichment of 30% is proposed to reduce the effect of the burn-up, and to improve the TBR. With an initial value for the enrichment of 30% in Li 6, the final value will be around 26%. Since the TBR is not very sensitive to this parameter, there will be no need for Li refuelling. Furthermore, the natural (or forced) circulation of the liquid breeder material will ensure a homogeneous distribution of the burn-up and of the enrichment within the blanket. The atomic composition of the eutectic at the end of INTOR's lifetime (6 MW y r / m 2) will be approximately

Lit6Pbs4. 3.2. Results The configuration and lay-out of the breeding blanket proposed is described in the previous section. It is the result of a trade-off study, in which not only the neutronic requirements have been taken into account, but also those coming from the thermohydraulic and mechanical analyses, the plasma-first wall interaction and material compatibility considerations. The first wall has an equivalent thickness of 23 ram, five of which are taken by the heavy water coolant. LiPb eutectic is contained in a tank with walls of 8 mm thickness. The coolant tubes have inner/outer diameters of 18/21 mm. The temperature of the coolant is in the interval 240-260 o C and its density is around 0.77 g/cm 3. The value of the TBR achievable with this blanket and the neutron balance are given in table 4. The heat deposition in the blanket is shown in fig. 5. The energy multiplication provided by this blanket is about 1.26; this means that the total energy deposited into the blanket by each fusion neutron is near to 17.5 MeV. Some improvement in the value of TBR may be obtained, by increasing the Li 6 enrichment. The calculations show that, with the enrichment at the level of 90%,

Table 3 Tritium breeding ratio versus thickness of the first wall

Table 4 Neutron balance of the breeding blanket

First wall thickness (ram)

Neutron multiplication Parasitic absorption

10

15 20 25

TBR (D20) 1.37 1.31 1.26 1.21

TBR (H20) 1.30 1.24 1.18 1.14

first wall includes the equivalent thickness of 4 mm of D20 or H20.

T h e

Neutron leakage Tritium breeding ratio

- first wall - breeder zone - from Li6 - from Li7 - Total

1.45 0.12 0.11 0.06 1.16 0.01 1.17

190

J. Airola et al. / A blanket concept with liquid Li tzPbsz

4. Thermomechanics

10 2 I: P l a s m a 2, 4, 6: AISI 316 3:D20 4: Homogeneous mixture of AISI 316, H 2 0 and LinPbe~ 10 ~

10-1

3

The thermomechanics analysis has been performed for the breeder unit configuration shown in figs. 1 and 2, by considering the most loaded one, at the central portion facing the plasma. Together with the reference design case (AISI-316 first wall acting as a radiation panel separated from the breeder unit), a second case has been analysed, namely: a configuration without a radiation panel. In that case the first wall coincides with the side wall of the breeder unit facing the plasma, the thickness of it remaining the same as in the first case (8 mm). For this second case, two analyses have been performed where the breeder unit vessel material is AISI-316 or ferritic steel. The calculations have been carried out under the following main assumptions: the vessel unit was considered to be in steady state, subject to the maximum (flat top) thermal loading shown in table 5. The surface heating is entirely absorbed by the radiation panel in the reference design; - inlet/outlet coolant temperatures: 240 and 260°C, respectively (at 5 MPa pressure); - heat transfer coefficient between cooling water and tube wall: 25 k W / m 2 ° C (corresponding to 3.3 m / s flow velocity); - constant thermal conductivity for liquid Li17Pbs3 breeder: 15.9 W / m °C; temperature dependent data for structural materials as shown in table 6; - perfect conductivity between steel and liquid Lil~ Pb83Two- and three-dimensional calculations of temperature and stress distributions were performed by the finite element method using the codes FLHE (Fullard [6]) and BERSAFE (Hellen [7]). -

10.2 I I I I ,00 .66 1.32 1,99 2.64

I

I I I I 3.97 4.63 5.29 5.95

Distance from first wall (cm) x 10

Fig. 5. Heat deposition in the blanket.

the TBR reaches the value of 1.22. This increase is the result of the decrease of the parasitic capture in the breeding region. An increase of the thickness of the breeding region by 6 cm could reduce the neutron leakage and allow a further gain in the TBR that would enrich the value of 1.24. The value of the TBR for the reactor will be much lower than the local values reported above, because the breeding blanket will be located only in the upper and outer regions of the torus, because in an experimental reactor many zones are occupied by test sections, and because of the numerous openings, windows and discontinuities. Goal of the "tritium supply reducing blanket" in INTOR-Phase-I was to produce at least 60% of the tritium needed to fuel the plasma. This goal appears to be achievable.

-

Table 5 Thermal loading in the breeder unit Thermal loading

With radiation panel

Without radiation panel

Surface heating (W/cm2)

No surface heating

12

Simple exponential decreases from 11.5 plasma side to 0.3 rear side Compound exponential decreases from 17.5 plasma side to 0.3 rear side

Simple exponential decreases from 14 plasma side to 0.4 rear side Simple exponential decreases from 14 plasma side to 0.4 rear side

Heat deposition (W/cm3) Vesselwall -

- Breeder

J. Airola et al. / A blanket concept with liquid LitTPba~

191

Table 6 Temperature dependent data for structural materials Material

AISI 316 Ferritic

Temp. ( o C)

Thermal conductivity K(W/m ° C)

Thermal expansion a ( 1 / o C)

Elastic modulus E(MPa)

Poisson's ratio

21 537

14.5 21.5

15.46 X 10.52 X

10 - 6

195 100 155 100

0.290 0.316

100 600

27.6 26.4

10.13 × 1 0 - 6 12.15 × 10 -6

190 000 150 000

0.290 0.316

4.1. Two-dimensional calculations

The two-dimensional temperature and stress calculations were performed for half of the breeder unit crosssection (due to structural and loading symmetry) (fig. 3) to obtain the temperature distributions in the LilTPbs3, in the wall of cooling tubes and in the wall of the vessel. An estimate of the stress distribution in the stainless steel wall of the vessel was also performed. A set of proper conditions has therefore been imposed to account for geometrical and loading symmetry. The above structure was analysed using 750 type-EP16 elements and 2700 nodes. The highest temperatures, equivalent Von Mises stresses and displacements are presented in table 7. Fig. 6 shows typical plots of temperature and Von Mises stress in the breeder and in the vessel wall for the design case with radiation panel. 4.2. Three-dimensional calculations

In these calculations one half (due to structural and loading symmetry) of the breeder unit has been considered assuming that it was free to expand both in the toroidal and in the poloidal directions according to real constraints. For the 3-D stress analysis, each cross-sec-

10 - 6

p

tion of the breeder unit was assumed to have the temperature distributions obtained in 2-D analysis. Indeed, it was shown that the variation of the cooling water temperature along the poloidai direction is very small, at maximum 5°C from top to bottom. The structure was analysed using 252 type EZ 60 elements and 2700 nodes. The highest temperature, Von Mises stresses and displacements are presented in table 8. Fig. 7 shows a plot of the deformed structure for the design case with radiation panel. 4.3. Parametric analysis

To assess the role played by the most important parameters as heat deposition, surface heating and stainless steel thermal characteristics on the maximum temperatures and stresses, use has been made of the Response Surface Methodology (RSM) techniques [810]. Input data for RSM have been taken from 2-D calculation results. The parameter ranges are as follows: - maximum heat deposition ( W / c m 3 ) l l . 5 surface heating ( W / c m 2) 0.0 - thermal conductivity K ( W / c m K) 0.15 - thermal expansion coefficient a ( 1 0 - 6 K -1) 10 -

+ 27, + 30, + 0.30,

Table 7 2-D max temperatures, Von Mises stresses and displacements

Temp. ( o C) - vessel wall - breeder Von Mises stress (vessel wall), MPa Displacement ( x / y ) (vessel wall), mm

Vessel in AISI 316 with radiation panel

Vessel in AISI 316 without radiation panel

Vessel in ferritic steel without radiation panel

367 354

488 433

458 418

32

100

42

2.2/-0.84

2.6/-0.9

1.5/-0.5

+22,

192

J. Airola et aL / A blanket concept with liquid Li lTPbaj 1.='~u-tt~

i.,~ I,D ~Di,.~

tot

t~r~'r~ ' ~ t i t r

t.n

to -

t"-

c~

t°

m

oa

t.n :o

t~

X

T

c~

m

N

laq

N 01

KEY (°C)

1 -

2 " ( "i 3 4 ~ 5 6 7

0.260000 0,279999 0.300000 0.320000 0.340000 0.350000 0.360000

E03 4" I E03 ~_1¢.,, E03C" ~ E03~-J E03 E03~,

1 0.260000 0.279999 3 0.300000 4 0.320000 5 0.340000 6 0.350000 7 0.360000 2

E030\

KEY (MPa)

E03 E03 E03 E03 E03 E03 E03

1 2 3 4 5

MAX 367 °C

AISt 316 MAX 367°C Lilt Pbe3MAX 354°C

0,328350 E01 0.955039 E01 0,158173 E02 0,220842 E02 0.283511E02

c~

MAX 32 MPa

0o00 Fig. 6. 2-D temperature and stress in the breeder unit with radiation panel.

Figures 8 to 10 show the behaviour of t e m p e r a t u r e a n d stress as functions of the selected parameters. 4.4. Results

T h e main results of the analysis can be summarized as follows: - F o r all cases (with a n d without radiation panel), the

highest equivalent Von Mises stress in 3-D calculations is a b o u t 3.5 times the highest value o b t a i n e d in 2-D calculations. This is due to the b e n d i n g m o m e n t which results from the n o n - u n i f o r m t e m p e r a t u r e distribution on each cross-section of the module. This b e n d i n g m o m e n t c a n n o t b e taken into account by 2-D calculations. - T h e reference design with radiation panel seems to be

Table 8 3-D max temperatures, Von Mises stresses and displacements

Temp.(°C)

- vessel wall - breeder

Von Mises stress (vessel wall), MPa Diplacernent ( x / y / z ) (vessel wall), mm

Vessel in AISI 316 with radiation panel

Vessel in AISI 316 without radiation panel

Vessel in ferritic steel without radiation panel

367 354

488 433

458 418

110

332

152

3/-0.8/-7.4

4.4/-2.4/-8.8

2.4/-1.2/-5.0

J. Airola et a L / A blanket concept with fiquid Li /TPbss

193 MAX SS HEAT DEP. = 11.5 w/cm = SURF. HEAT. --~ 12 w/cm ~

SURF. HEAT. = 12 w/cm J

~ 600

600 500

500

IE

N 400

11.5 15

400 N 300 L

2o

25 27

~ *-

J300

0.1s 0.20 0.25 0.30

A) MAX SS HEAT DEPOSITION Iw/cm=l

B) SS THERMAL COND. Iw/cmOC)

sootMAXSS HEAT DEP. --~ 11.5 w/crn3

5

~

]600

~

500

w 350

:, I

0

5 10 15 20 C) SURFACE HEATING Iw/cm~)

350 25

Fig. 9. Parametric analysis: LilTPbs3 maximum temperature.

Fig. 7. 3-D undeformed and deformed configuration of the breeder unit with radiation panel.

feasible. The ASME code [11] for AISI 316 at 367 ° C and with alternating stress from 0 to 110 MPa allow more than 106 cycles. This value exceeds the estimated INTOR-Phase-I lifetime (7 x 105 cycles). Moreover with the arrangement and size of the cooling tubes shown in figs. 2 and 3, the maximum temperature of Li]TPb83 is within the prescribed AISI 316 stainless steel compatibility limit of 400°C.

700t SURF" HEAT. =

MAX SS HEAT DEP. = 11.5w/cm3

12w/~

600~URF.HEAT,=12 w/cm} 6OO

700

5CO~

~ 500

- The design in AISI 316 without radiation panel with 488 ° C as maximum temperature and an alternating stress from 0 to 332 MPa can stand about 3 x 103 cycles (ASME code [11]). The maximum temperature of Li]vPbs3 (433°C) slightly exceeds the AISI 316 compatibility limits, even if restricted to a small zone facing the plasma. - The design with ferritic steel with 4 5 8 ° C as maximum temperature and an alternative stress from 0 to 152 MPa can stand more than 106 cycles (ASME code [11]). The maximum temperature in LilvPb83 (418 o C) possibly is acceptable from the compatibility point of view even if there are only a few experimental data on ferritic steels. - From the results of parametric 2-D analysis by RSM

500

~130 SURF.HEAT.= 12 w / c m

MAX SS HEATOEP = 11.5w/cm= SURF.HEAT.= 12 w/cruz

~

500

11.5 15

20

25 27

A) MAX SS HEAT DEPOSITION (w/cm 3)

~eoo

0.15 0.20 0.25 0.30 B) SS THERMAL COND. (w/cm°CI

t MAX SS HEAT DEP. = 1

500

35O0

1

~

~

600

11,5 15

25 27

'30f~'30 ~10o

015

0.20

0.25

0.30

~'°°~" ~ ~.,l'°m' °

B) SSTHERMALCONO.(w/cmIC)

MAX SS HEATOEP.= 11,5w/r.m~

MAX S5 HEAT DEP.= 11.5wlcm)

SURF.HE. . . . .

1oo

500

0

5 ~0 15 20 25 C) SURFACE HEATING (w/cm a)

20

A) MAX SS HEATDEPOSITIONIw/cmj)

30

Fig. 8. Parametric analysis: SS maximum temperature.

5

10

15

20

C) SURFACEHEATINGIw/cm~)

25

10

15

20

25

O) THERMALEXPANSIONCOEFF.(lO~/°CI

Fig. 10. Parametric analysis: SS maximum stress (2-D calculations).

194

J. Airola et al. / A blanket concept with liquid LilzPb8~

it appears that stress and temperature behaviour are almost linear as functions of the selected parameters (thermal loads and material characteristics). To obtain the maximum stress in 3-D geometry it is sufficient to multiply the results of 2-D analysis by a factor 3.5.

5. Conclusions The conceptual design of a tritium producing blanket for I N T O R / N E T using liquid LilTPbs3 as a breeder material has enabled us to derive some conclusions, even if the experimental data base for this material is still not complete. They are: - A design with the first wall separated from the breeder unit looks feasible from the point of view of temperatures and stresses for normal operation conditions. In particular the maximum temperature in the breeder is sufficiently low to avoid serious corrosion problems. - The local breeding ratio which can be achieved with the compact configuration here proposed makes it attractive not only for the tritium breeding requirement of I N T O R but also in view of obtaining a self-sufficient breeding in power reactors. - The problems of keeping liquid the breeder during operation can be handled without major difficulties. - The consequences on the breeder unit vessel of a coolant tube rupture appears severe and deserves more refined analysis and experimental verification. Studies are in progress at J R C - I s p r a to investigate blanket designs with cylindrical vessel units arranged along the poloidal or toroidal direction. Such solutions appear more attractive from the safety point of view, in

case of coolant-tube rupture, but less effective for tritium breeding. References [1] F. Carrr, Z. Tilliette and J. Remoleur, Conceptual study of a lithium-lead eutectic blanket for a power reactor, 5th Topic Meeting on the Technology of Fusion Energy, Knoxville, April 26-28, 1981. [2] M. Abdou et al., Demonstration Tokamak power plant, Fusion Technology 1 (1982) 617-618, 12th SOFT Conf. Jialich, FRG, September 1982. [3] B. Badger et al., WITAMIR-I, a university of Wisconsin tandem mirror reactor design, Univ. of Wisconsin report, UWFDM-400 Ch. VIII (September 1980). [4] INTOR-Phase-I, IAEA Vienna (1982) Panel Proc. Series. [5] INTOR-Phase-IIA, Critical Issues, Eur. Contr. to the INTOR-Phase-II Workshop, 1982, EUR FU BR/XII132/82/EDV30, Ch. VIII, pp. 112-123. [6] K. Fullard, FLHE Phase 2, Central Electricity Generating Board, Research Dept., Berkeley Nuclear Laboratories, Rep. RD/B/N286 (January 1974). [7] T.K. Hellen, Bersafe Phase II (Level 1), Central Electricity Generating Board, Research Depat., Berkeley Nuclear Laboratories, Rep. RD/B/M2508 (May 1974). [8] G.E.P. Box and J.S. Hunter, Multifactor experimental designs for exploring response surface, Ann. Math. Statist. 28 (1957) 195. [9] B. Lisanti, Una applicazione della RSM all'analisi sperimentale di un LOCA in un PWR, Doctorate Thesis, Pisa (Italy University, 1980, unpublished). [10] L. Olivi, RSM Handbook for Nuclear Safety Analysis, Part 2, Program Library, Tech. Note No. 1.05.01.82.56 (May 1982) Commission of the European Communities, JRC~Ispra Establishment, Italy. [11] ASME-boiler and pressure vessels. Code case No. 47, Elastic analysis.