neutronic and photonic analysis of the water-cooled pb17li test blanket module for iter-feat

Neutronic and photonic analysis of the water-cooled Pb�/17Litest blanket module for ITER-FEAT

G. Vella a,�, P. Chiovaro a, P.A. Di Maio a, A. Li Puma b, E. Oliveri a

a Dipartimento di Ingegneria Nucleare, Universita di Palermo, Viale delle Scienze, 90128 Palermo, Italyb CEA-Saclay, DEN/DM2S/SERMA/LCA, 91191 Gif-sur-Yvette, France

Abstract

Within the European Fusion Technology Program, the Water-Cooled Lithium Lead (WCLL) DEMO breeding

blanket line was selected in 1995 as one of the two EU lines to be developed in the next decade, in particular with the

aim of manufacturing a Test Blanket Module (TBM) to be implemented in ITER. This specific goal has been

maintained also in ITER-FEAT program even if the general design parameters of the TBMs have reported some

changes. This paper is focused on the investigation of the WCLL-TBM nuclear response in ITER-FEAT through

detailed 3D-Monte Carlo neutronic and photonic analyses. A 3D heterogeneous model of the most recent design of the

WCLL-TBM has been set-up simulating realistically its new lay out and taking into account 9% Cr martensitic steel as

structural material. It has been inserted into an existing 3D semi-heterogeneous ITER-FEAT model accounting for a

proper D�/T neutron source. The analyses have been performed by means of MCNP-4C code running on a cluster of

four workstations through the implementation of a parallel virtual machine. The main WCLL-TBM nuclear responses

have been determined focusing the attention on power deposition density, material damage through displacement per

atom (DPA) and He and H production rate, daily tritium production and tritium production rate radial distribution in

the module. Moreover, the impact of using lithium at various Li6 enrichment on the TBM nuclear response has been

investigated. The results obtained are herewith presented and critically discussed.

# 2002 Published by Elsevier Science B.V.

Keywords: Breeding blanket; Neutronic; Photonic; Monte Carlo method

1. Introduction

The Water-Cooled Lithium Lead (WCLL) blan-

ket is one of the two European blanket lines

selected by the European Blanket Selection Ex-

ercise for a future DEMOnstration reactor [1]. The

European Fusion Technology Program has fore-

seen intense research activities on this topic and,

among them, the design and manufacturing of a

WCLL Test Blanket Module (WCLL-TBM), to be

tested in ITER-FEAT, seems to be the most

attractive one as its overall objective relies in the

evaluation of the nuclear, thermal�/hydraulic and

thermal�/mechanical behavior of the WCLL blan-

ket under conditions which are, as much as

possible, relevant to DEMO [2].

� Corresponding author. Tel.: �/39-091-232-250; fax: �/39-

091-232-215

E-mail address: [email protected] (G. Vella).

Fusion Engineering and Design 61�/62 (2002) 439�/447

www.elsevier.com/locate/fusengdes

0920-3796/02/$ - see front matter # 2002 Published by Elsevier Science B.V.

PII: S 0 9 2 0 - 3 7 9 6 ( 0 2 ) 0 0 2 2 6 - 0

mailto:[email protected]

Within the framework of WCLL-TBM designactivities, a cooperation between the Department

of Nuclear Engineering (DIN) of the University of

Palermo and the Commissariat a l’Energie Ato-

mique (CEA) has started with the specific aim of

investigating the nuclear response of the afore-

mentioned module under ITER-FEAT working

conditions, finding out, at the same time, the

proper information needed to study its thermal�/

hydraulic and thermal�/mechanical behavior.

A detailed 3D nuclear study of the most recent

design of the WCLL-TBM has been performed

and the results obtained are herewith presented

and critically discussed.

2. Outline of the WCLL-TBM

The WCLL-TBM aims to represent the equa-

torial part of an inboard segment of the corre-

sponding DEMO blanket and it is expected to use

Pb�/17Li liquid eutectic alloy as breeder and

neutron multiplier material, sub-cooled pressur-

ized light water, under typical PWR conditions, as

coolant and reduced-activation martensitic steel as

structural material [2].The WCLL-TBM should be placed in one of

ITER-FEAT equatorial ports since from the

reactor first day operation till to the end of its

Basic Performance Phase. In particular, it has to

be cased in one vertical half of an ITER-FEAT

mid-plane port, being housed in a water-cooled

steel frame directly supported by the vacuum

vessel. Moreover, it has a poloidal height of 1.72m, a toroidal width of 0.514 m and a radial depth

of 0.585 m [2].

From the structural point of view, the afore-

mentioned module has a box-shaped structure

composed of a Segment Box (SB), a Breeder

Zone (BZ) and a cooling system.

The SB is a directly-cooled steel box having,

basically, the function of Pb�/17Li container. It ismainly composed of a First Wall (FW), two Side

Walls (SWs) and a Back Plate (BP). Moreover, it is

reinforced by toroidal and radial steel stiffeners,

which are intended to withstand the disruption-

induced forces and the full water pressure under

faulted conditions. The reference structural mate-

rial is reduced-activation 9% Cr martensitic steel,called EUROFER, similar to the commercial one

code-named Z10 CDVNb 91 [2].

The BZ, subdivided by the stiffeners in 12

poloidal sectors, houses the Pb�/17Li liquid eu-

tectic alloy (90% Li6 enriched) which is slowly

circulated from the top to the bottom of the

module, letting tritium extraction, purification

and Li adjustment to take place in specific unitsoutside the blanket [3].

The cooling system is articulated in two inde-

pendent circuits, one cooling the SB and one

cooling the BZ. The former is composed of 65

toroidal�/radial cooling tubes and two vertical

manifolds welded behind the BP, while the latter

is composed of a bundle of 35 double walled C-

shaped tubes (DWTs) maintained in place withone spacer grid located at about TBM mid-plane

level. Two coolant headers are foreseen for the BZ

cooling circuit, one for the inlet and the other for

the outlet water, placed, respectively, in the

bottom and in the top part of the module.

The SB cooling circuit uses tubes with inside/

outside diameters of 8/10 mm, while the BZ one is

composed of double walled tubes having inside/outside diameters of 11/13.4 mm for the inner one

and 13.6/16.5 mm for the outer one. A thin copper

interlayer, 100 mm thick, is adopted in both the SB

and BZ tubes as brazing material and tritium

permeation barriers are adopted in the last ones to

reduce tritium release in the coolant under the

safety-prescribed limits.

The coolant is subcooled water at pressure of15.5 MPa, whose inlet and outlet temperatures are

foreseen to be 315/325 and 303/325 8C, respec-

tively, for the BZ and SB circuits [2].

3. Models and results

A detailed 3D study of the WCLL-TBM nuclear

behavior has been performed with the MonteCarlo method, using the Monte Carlo N-Particle

(MCNP) code ver. 4C and adopting the FENDL2

transport cross section library [4,5]. In order to

speed up calculations and owing to the improved

multiprocessing capability of the MCNP code, the

analyses have been carried out on a cluster of four

G. Vella et al. / Fusion Engineering and Design 61�/62 (2002) 439�/447440

workstations with heterogeneous operating sys-tems by adopting the PARALLEL VIRTUAL MA-

CHINE (PVM) software.

3.1. The model

In order to perform the aforementioned study a

pre-existing 3D model of ITER-FEAT, realisti-

cally simulating a blanket sector, has been

adopted. It represents 1/18 of the whole reactortoroidal extension (208) and it is delimited by two

proper reflecting surfaces located at toroidal

boundaries. The model describes in detail the

shielding blanket, the divertor cassette, the magnet

system, the vacuum vessel with its three major

ports and the cryostat. A proper D�/T neutron

source has been taken into account too.

A detailed 3D heterogeneous model of theWCLL-TBM with its steel frame has been set-up

and it has been inserted into the equatorial port.

The WCLL-TBM has been modeled in a fully

heterogeneous way, as appears from Fig. 1, with

few simplifications concerning the BZ tubes and

their collectors. In fact, the BZ cooling tubes have

been modeled as straight tubes without taking intoaccount their typical C shape and their collectors.

A particular attention has been paid to the

modeling of both SB and BZ tubes internal

structure, where a thin copper layer, 100 mm thick,

has been included to simulate the presence of the

brazing material.

The steel frame has been modeled as a homo-

geneous mixture of water and martensitic steel.Proper materials have been taken into account and

their macroscopic densities in operating conditions

are summarized in Table 1.

3.2. The results

The WCLL-TBM nuclear response has been

investigated, paying particular attention to neu-

tronic and photonic deposited power, tritium

production, helium and hydrogen production

rate and displacement per atom (DPA) in the

structural material.

The analyses have been carried out simulating alarge number of histories (10 000 000) and the

obtained results are affected by statistical uncer-

tainties lower than 3%.

3.2.1. Neutronic and photonic power deposition

The total power deposition in the module has

been evaluated in order to provide useful data for

the investigation of the WCLL-TBM thermal�/

hydraulic performances. It has been estimated

that the total power deposited by neutrons and

photons in the whole module is about 512 kW and

a detailed description of power deposition distri-

bution has been obtained. A summary of theseresults is reported in Table 2.

The deposited power seems to be decreased by

about a factor 2 with respect to the one concerning

the TBM previous design, conceived to be housed

in ITER-FDR [3], mainly due to the fact that the

Fig. 1. The WCLL-TBM model (midplane toroidal�/radial

section).

Table 1

Material macroscopic densities (kg/m3)

9% Cr steel 7760

Pb�/17Li 9510

Cu 8830

Water 727

G. Vella et al. / Fusion Engineering and Design 61�/62 (2002) 439�/447 441

ITER-FEAT Neutron Wall Load (NWL) (0.78

MW/m2) is 0.62 times the ITER-FDR one.

In order to find out useful information for the

WCLL-TBM thermal�/mechanical study, a de-

tailed nuclear analysis has been performed aiming

to obtain the radial distribution of the deposited

nuclear power density both in the SB and BZ.

Since the variations expected for the deposited

power density in both toroidal and poloidal

directions are negligible with respect to the ones

foreseen along the radial direction, they have not

been investigated to reduce calculation time.

The radial profiles obtained are shown in Figs.

2�/5. As it was expected the deposited power

density reaches its maximum value near theplasma-facing region both in the SB and BZ,

decreasing in a quite exponential way along the

radial direction.

The radial profiles of the power density depos-

ited in the Pb�/17Li eutectic alloy, in the structural

material (SS) and water (H2O) of the SB and in the

top (TC) and bottom (BC) caps have been fitted

throughout the following functions:

q§Pb�17Li(r)�11:495 exp(�0:3866r)

�3:178 exp(�0:0732r) (1)

q§SS�SB(r)�3:286 exp(�0:1645r)

�2:7303 exp(�0:0616r) (2)

q§H2O�SB(r)�3:722 exp(�0:0799r) (3)

q§TC(r)�3:591 exp(�0:1978r)

�2:403 exp(�0:0694r) (4)

q§BC(r)�3:209 exp(�0:272r)

�3:607 exp(�0:0804r) (5)

where r represents the radial distance from the

TBM plasma-facing surface and it is expressed in

cm. The minimum correlation factor obtained in

the curve fitting is higher than 0.998.

Table 2

WCLL-TBM power deposition distribution (kW)

WCLL-TBM 511.9

Breeder Zone 333 Segment Box 178.9

Pb-17Li 303.9 First wall 116.9

DWTs 7 Side walls 45.7

Water 10.2 Back plate 1.3

Stiffners 11.9 Top cap 7.2

Bottom cap 7.8

Fig. 2. Power density distribution in the WCLL-TBM Pb�/17Li eutectic alloy.


Fig. 3. Power density distribution in the toroidal rows of DWTs of the WCLL-TBM.

Fig. 4. Power density distribution in the WCLL-TBM SB.


The best fitting functions have been reported inFigs. 2�/5 in comparison with code predictions and

they are to be considered valid only in the range

2.65/r 5/55.50.

3.2.2. Tritium production

As one of the WCLL-TBM main goal is to test

the tritium breeding recovery and confinement

capability of such a kind of blanket line, the daily

tritium production together with the tritium pro-

duction rate density radial distribution have beeninvestigated.

The daily tritium production depends, ob-

viously, on the lithium enrichment of the Pb�/

17Li eutectic alloy and on the ITER-FEAT duty

cycle. Assuming a 90% Li6 enrichment and a duty

factor of 0.22, the daily tritium production has

been calculated to be about 14.9 mg/day, which is

in good agreement with the value calculated in [3]for the previous WCLL-TBM design, as far as the

different NWLs and duty cycles of ITER-FEAT

and ITER-FDR are considered.

The impact of Li6 enrichment on the daily

tritium production has been investigated too and

the main results are reported in Fig. 6. The results

obtained seem to suggest that the increase of Li6

enrichment has a strong impact on the daily

tritium production till the enrichment is lower

than 60%, while that impact decreases when the

enrichment becomes higher than 60%. In fact,

when the Li6 enrichment rises up from the natural

7.5 to 60% the daily tritium production substan-

tially doubles going from about 6 mg/day to a

value of 13 mg/day, while when it goes from 60 to90% a change of about 2 mg/day is predicted.

In order to provide useful data for tritium

permeation analyses it has been determined the

radial distribution of the tritium production rate

density, showed in Fig. 7.

3.2.3. Radiation damage

During plant operation, the highly energetic

fusion neutrons coming out from plasma continu-

ously interact with the structural material atomsinducing two main damage mechanisms for the

aforementioned materials.

The first mechanism is due to the displacement

of the atoms from their lattice positions as

consequence of collisions, while the second is

determined by the gas production as result of

Fig. 5. Power density distribution in the Caps of the WCLL-TBM.


various nuclear reactions mainly of (n, p) and (n,

a) kind. While the hydrogen isotopes diffuse out of

the metallic lattice or form metal hydrides, a-

particles remain trapped in the metal and generate

helium bubbles. These processes lead to unfavor-

able changes of mechanical properties (such as

embrittlement), limit the lifetime of the structural

material and affect their reweldability [3].

Fig. 6. Daily tritium production vs. lithium enrichment in Li6.

Fig. 7. Tritium, helium and hydrogen production rate radial distributions.


In order to investigate the level of WCLL-TBMmaterial damage due to the first mechanism it has

been evaluated the DPA adopting the displace-

ment cross sections for iron taken from ASTM

standards [6]. Supposing a 0.22 duty factor reactor

operation during the whole year, it has been

evaluated the DPA distribution along the radial

depth of the WCLL-TBM SB structural material

showed in Fig. 8.In order to estimate the effect of the second

damage mechanism, He and H production rate

distributions have been evaluated too, along the

radial depth of the SB structural material. Fig. 7

reports the profiles obtained and it can be deduced

that, as a consequence of the ITER-FEAT reduced

NWL and duty factor, their maximum values are

lower (about one third) than the ones reported in[3].

4. Conclusions

A detailed investigation of the WCLL-TBM

nuclear response in ITER-FEAT has been per-

formed by means of 3D-Monte Carlo neutronic

and photonic analyses.

A 3D heterogeneous model of the new version

of the WCLL-TBM has been inserted into an

existing ITER-FEAT model accounting for a

proper D�/T neutron source and reduced-activa-

tion 9% Cr martensitic steel has been adopted as

structural material.

The main WCLL-TBM nuclear responses have

been determined, such as detailed power deposi-

tion density, material damage through DPA and

He and H gas production rate, radial distribution

of tritium production rate density and daily

tritium production in the module. Moreover, it

has been investigated the impact of using lithium

at various Li6 enrichment on the TBM breeding

performances.

The nuclear power deposition in the TBM has

been estimated to be 512 kW and its detailed

distribution has been evaluated too. A clear

reduction in the values of the aforementioned

variables has been found with respect to the

previous design, mainly due to the reduced

ITER-FEAT NWL.

Fig. 8. DPA per year in the WCLL-TBM SB.


As far as the radiation damage is concerned,both the DPA and He and H production rate

distributions have been calculated along the radial

depth of the TBM structural material, highlighting

that their maxima are, obviously, achieved in the

FW proximity where the neutron fluence is higher.

The maxima achieved at the end of 1 year of 0.22

duty factor operation have been estimated to be

about 11.8 and 35 appm, respectively, for He andH and about 1.05 for the DPA.

The daily tritium production together with its

production rate radial distribution have been

evaluated, observing that the former achieves a

value of about 14.9 mg/day.

At the same time the influence of the Li6

enrichment on the TBM nuclear response has

been evaluated, enabling to conclude that eutecticalloy enrichment in Li6 higher than 60% is

mandatory for a significant tritium breeding to

take place in the TBM, but highlighting also that it

can be optimized since a 90% Li6 enriched breeder

does not seem to be a first order necessity for the

WCLL-TBM.

A further improvement of the study could be

achieved modeling the real structure of the C-

shaped tubes and investigating the potential effectof the employment of innovative structural mate-

rials.

References

[1] L. Giancarli, G. Benamati, M. Futterer, G. Marbach, C.

Nardi, J. Reimann, Development of the EU water-cooled

Pb�/17Li blanket, Fusion Engineering and Design 39�/40

(1998) 639�/644.

[2] Y. Poitevin, L. Giancarli, A. Li Puma, J.-F. Salavy, J.

Szczepanski, Status of the design and testing programme of

the WCLL Test Blanket Module for ITER-FEAT, Rapport

SERMA/RT/01-3019/A.

[3] G. Vella, G. Aiello, M.A. Futterer, L. Giancarli, E. Oliveri,

F. Tavassoli, Water-cooled Pb�/17Li test blanket module

for ITER: impact of the structural material grade on the

neutronic responses, Journal of Nuclear Materials 258�/263

(1998) 357�/361.

[4] J.F. Briesmeister, MCNP, A General Monte Carlo N-

particle transport code. LA-12625-M, Version 4C (2000).

[5] M. Herman, A.B. Pashchenko (Eds.), Extension and

Improvement of the FENDL Library for Fusion Applica-

tions (FENDL-2), Report on a Meeting Held in Vienna,

Austria, 3�/7 March 1997, Report INDC (NDS)-373, 1997.

[6] Annual Book of ASTM standards, vol. 12.02 Nuclear (II)

Solar and Geothermal Energy, E693�/E694, 1994.


neutronic and photonic analysis of the water-cooled pb17li test blanket module for iter-feat

Documents