The objective of the Spanish national project TECNO_FUS is to generate a conceptual design of a DCLL (Dual-Coolant Lithium-Lead) blanket for the DEMO fusion reactor. The dually-cooled breeding zone is composed of He/Pb-15.7 6Li and SiC as liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer-97) for the blanket and austenitic steel (316LN) for the Vacuum Vessel (VV). The goal of this work is to analyze the radioactive waste production by the neutron-induced activation and the back-end of the blanket and the VV (SS316LN) materials (Eurofer, SiC, LiPb, and SS316LN). Furthermore, the radioactive waste production in the cryostat (SS316LN) and the bioshielding (concrete) has been estimated. Following the current approach to the back-end of the materials in fusion facilities, the radioactive waste has been subdivided according to the activity-level classification (EW, exempted waste, LILW, low and intermediate level waste, and HLW, high level waste) and according to the radiological complexity of operations (handling and cooling). The activation calculations have been carried out with the ACAB code

Catalán, J.P.

Ogando Serrano, Francisco

Sanz Gonzalo, Javier

Archivo Digital UPM

Activation Analysis for a He/LiPb dual Coolant Blanket for DEMO Reactor

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ACTIVATION ANALYSIS FOR A He/LiPb DUAL COOLANT BLANKET FOR DEMO REACTOR  
 
 
J. P. Catalán1, F. Ogando1,2, J. Sanz1,2 
 
1 Departamento de Ingeniería Energética, UNED, C/ Juan del Rosal 12, 28040 Madrid, SPAIN, jpcatalan@bec.uned.es  
2  Instituto de Fusión Nuclear, UPM, Madrid, SPAIN 
 
 
The objective of the Spanish national project 
TECNO_FUS is to generate a conceptual design of a 
DCLL (Dual-Coolant Lithium-Lead) blanket for the 
DEMO fusion reactor. The dually-cooled breeding zone is 
composed of He/Pb-15.7 6Li and SiC as liquid metal flow 
channel inserts. Structural materials are ferritic-
martensitic steel (Eurofer-97) for the blanket and 
austenitic steel (316LN) for the Vacuum Vessel (VV). The 
goal of this work is to analyze the radioactive waste 
production by the neutron-induced activation and the 
back-end of the blanket and the VV (SS316LN) materials 
(Eurofer, SiC, LiPb, and SS316LN). Furthermore, the 
radioactive waste production in the cryostat (SS316LN) 
and the bioshielding (concrete) has been estimated. 
Following the current approach to the back-end of the 
materials in fusion facilities, the radioactive waste has 
been subdivided according to the activity-level 
classification (EW, exempted waste, LILW, low and 
intermediate level waste, and HLW, high level waste) and 
according to the radiological complexity of operations 
(handling and cooling). The activation calculations have 
been carried out with the ACAB code. 
 
I. INTRODUCTION 
 
The framework of this study is the Spanish national 
project TECNO_FUS, whose objective is to carry out a 
design of a Dual-Coolant Lithium-Lead (DCLL) blanket 
for a DEMO fusion reactor. Particularly, in this work the 
neutron-induced activation of the reactor materials is 
analyzed from the point of view of the waste 
management.   
The blanket design is based on the C model of the 
European Fusion Power Plant Conceptual Study (PPCS1) 
with modifications to reactor parameters and plant 
systems specifications. The neutron wall loading (NWL) 
average is 2.1 MW/m2 for the blanket first wall with 3450 
MW of fusion power. 
The blanket has a dually-cooled breeding zone with 
Pb-15.7 6Li (90% 6Li enrichment) serving as breeder and 
coolant, and with pressurized helium as primary coolant. 
The reduced activation ferritic-martensitic (RAFM) steel, 
Eurofer-97, is used as structural material in the blanket 
and 316-LN austenitic steel in the vacuum vessel (VV). 
The flow channel inserts of the LiPb liquid metal are 
made of SiC serving as electrical and thermal insulator. 
Following the current approach to the back-end of the 
fusion materials,2 the activation calculations have been 
performed to obtain contact dose rate and decay heat. 
These quantities allow us to categorize the different 
materials according to the radiological complexity of 
operation and the activity-level classifications. Also the 
activity limit for the recycling in foundries and the 
clearance index has been taken into account for the 
cryostat and the bioshielding where these limits can be 
fulfilled.  
The activation calculations have been carried out 
with the ACAB3 code with EAF2007 libraries. The 
neutron fluxes from MCNPX have been processed with 
the MC2ACAB4 code (modified version of 
TARTREAD5) that generates the ACAB input 
automatically.  
 
II. METHODOLOGY 
 
II.A. Neutronic Calculation and Geometry Model.  
 
The neutronic model6 (Fig. 1) is a simplified layered 
model generated with an Excel/CATIA interface and 
transferred to MCNP geometry format by MCAM.7 
Actually the divertor and coil materials of the design are 
not accurately enough defined and are not considered in 
the activation calculations. 
 
 
Fig. 1. Geometry of the MCNP neutronic model. 
738 FUSION SCIENCE AND TECHNOLOGY        VOL. 60        AUG. 2011
II.B. Activation Calculations.  
II.B.1. Irradiation Scenarios. 
In the activation calculations two irradiation 
scenarios have been assumed: i) a DEMO scenario with 
limited irradiation time, ii) a power plant scenario 
considering its lifetime: 
 DEMO scenario: 20000 hours of continuous 
irradiation. 
 Power Plant scenario:  
Ƈ  5 years of continuous irradiation for the blanket 
components (structure, Flow Channel Inserts 
and liquid metal) and 8 replacements during 
plant life time.  
Ƈ 40 years of continuous irradiation for the 
permanent components (Vacuum Vessel, 
Cryostat and Bioshielding). 
II.B.2. Composition of the Materials. 
The compositions of the materials (with impurities) 
Eurofer (specified composition), SiC and SS316LN are in 
Ref. 8-10. The LiPb used is Pb-15.7 6Li (using lithium 
enriched to 90 % 6Li) with the impurities listed in Table I: 
 
TABLE I. LiPb Impurities. 
Element ppm 
Al 1 
P 10 
Ti 10 
Cr 3 
Fe 13 
Ni 7 
Cu 1 
Ga 1 
As 40 
Ag 1 
Cd 1 
In 1 
Sb 10 
Au 1 
Hg 1 
Bi 200 
II.C. Waste Management Criteria.  
II.C.1. Radiological Complexity of Operation. 
The actual way to analyze the radiological 
complexity of radioactive waste operation2 takes into 
account two factors: handling and cooling. The handling 
limits are based on the dose limits that a worker or a 
machine can tolerate while handling the waste, and the 
quantity used for these limits is the contact dose rate (CD, 
in this work the semi-infinite slab approximation is 
assumed):  
 Hands-on Handling (HOH): CD ȝ6YK.  
 Shielded Hands-on Handling (SHOH)  ȝ6YK 
CD < 2mSvh. 
 Remote Handling (RH): CD> 2mSv/h. 
The Cooling limits are based on the cooling 
necessary for the safe storage of radioactive waste and the 
quantity of decay heat (DH): 
 No-active cooling: DH* < 10 W/m3. 
 Dry cooling: 10 W/m3 < DH < 2 kW/m3. 
 Wet cooling: DH > 2 kW/m3. 
The radiological complexity of operation can be 
divided in different levels (Table II) assigning different 
values to the previous criteria and summing them up. 
 
TABLE II. Radiological complexity of operation. 
Handling (H) Cooling (C) Difficulty  score (H+C) 
HOH=1 NONE=0 level 1 1 
SHOH=2 DRY=3  level 2 2 
RH=3 WET=5 level 3 3,4,5 
    level 4 6,7,8 
 
Level 1 materials can be handled hands-on and no 
cooling is required. Level 2 materials can be handled 
using shielded hands-on methods and again no cooling is 
required. In level 3 materials require remote handling 
and/or dry cooling. Finally, level 4 materials require 
active cooling and no operation of material is possible.  
 
II.C.2. Activity-level Classification.  
 
The activity-level classification is taken from 
IAEA.11 In our case we use a classification based only on 
the clearance index and the decay heat. 
 Exempted Waste (EW): below the Clearance limit. 
 Low and intermediate level waste (LILW): DH < 2 
kW/m3.  
 High level waste (HLW): DH > 2 kW/m3.  
 
III. WASTE ASSESSMENT  
 
III. A. Blanket and Vacuum Vessel.  
 
In order to classify the radioactive waste within the 
above mentioned criteria, the contact dose and the decay 
heat (rates) have been analyzed for the different materials 
of the blanket and the VV. The critical isotopes for these 
quantities have been also identified at different cooling 
times. 
In the following subsections the most critical zones 
for each material are analyzed. These zones are the first 
layers (in radial direction) of Eurofer (blanket first wall, 
FW), SiC (breeding zone), LiPb (breeding zone) and 
SS316LN (inner wall of the VV).  
FUSION SCIENCE AND TECHNOLOGY        VOL. 60        AUG. 2011 739
Catalán et al.        ACTIVATION ANALYSIS OF DCLL BLANKET FOR DEMO REACTOR
III.A.1. Eurofer. 
 
For the first wall of Eurofer the contact dose meets 
the SHOH limit before 100 years of cooling time but does 
not meet the HOH limit in a reasonable time (more than 
1000 years). The dry cooling limit is fulfilled after less 
than five years and active cooling is not needed after the 
time interval between 10 and 50 years. 
 
 
Fig. 2. Contact Dose and Decay Heat for the first wall of 
Eurofer under DEMO scenario.  
 
The critical isotopes for the HOH limit are Co 60 up 
to 100 years and Nb 94 after this time. The latter does not 
allow us to fulfill the HOH limit in a reasonable time 
(even after 1000 years). 
 
III.A.2. SiC. 
 
Commercial SiC (Fig. 3) meets the SHOH limit 
before 50 years of cooling time and always is in the no-
active cooling regime.   
The critical isotopes for the HOH limit are Co 60 up 
to 100 years and Nb 94 after this time. Like in 
EUROFER, Nb 94 does not allow us to fulfill the HOH 
limit in reasonable time (even after 1000 years).  
 
III.A.3. Lithium-Lead. 
 
The contact dose (Fig. 4) does not meet the SHOH 
limit in a reasonable time (even after 1000 years). The 
decay heat is always in the dry cooling zone  before 100 
years of cooling time, while after that period of time it is 
in the no±active cooling zone. 
  The critical isotopes for the SHOH limit are Bi 207 
up to 100 years and Bi 208 after this time. The latter is the 
most critical because it does not allow meeting the SHOH 
in a reasonable time (even after 1000 years). 
 
 
Fig. 3. Upper: Contact dose of the first layer of SiC 
under DEMO scenario. Lower: Decay heat for the first 
layer of SiC under DEMO scenario. 
 
 
Fig. 4. Contact Dose and Decay Heat for the first 
layer of lithium-lead under DEMO scenario. 
 
III.A.4. SS316LN. 
 
For DEMO irradiation scenario simulated for the 
SS316LN steel located in the inner wall of the VV (Fig. 5 
and 6) the SHOH limit is fulfilled after around 100 years 
of cooling time, the no-active cooling requirement is 
achieved after 50 years (Fig. 6). The value of decay heat 
740 FUSION SCIENCE AND TECHNOLOGY        VOL. 60        AUG. 2011
Catalán et al.        ACTIVATION ANALYSIS OF DCLL BLANKET FOR DEMO REACTOR
in the steel is always below the dry cooling limit. For 
plant scenario (Fig. 7) the SHOH is not fulfilled in a 
reasonable time (even after 1000 years) while the no-
active cooling is fulfilled around 50 years. 
  
 
Fig. 5. DH and CD in the VV inner wall under DEMO 
scenario. 
 
Fig. 6. The Vacuum Vessel inner wall in both scenarios 
DEMO and Power Plant. 
 
For the contact dose, the most critical isotopes are Co 
60 up to 100 years and Nb 94 beyond this time. For the 
decay heat, they are Co 60 up to 100 years, Ni 63 from 
100 years to 400-500 years and Nb 94 beyond this time. 
 
III.B. Cryostat and Bioshielding. 
 
The cryostat and the bioshielding have been analyzed 
separately because they can be assessed using more 
aggressive waste management options (recycling in 
foundries and clearence) than for the blanket and VV 
components. The cryostat is made of SS316LN and 
fulfills the limit of 1000 Bq/g for recycle in foundries 
within 100 years of cooling time (10-50 years for DEMO 
scenario and at 100 years for power plant scenario, Fig.8). 
 
 
Fig. 7. Upper: clearance in the bioshielding. Lower: 
activity in the cryostat.  
 
The bioshielding is made of conventional concrete 
with 50 cm of thickness and fulfills the clearance level 
before the first year in both irradiation scenarios (Fig. 8).   
 
III.C. Management. 
 
III.C.1. Radiological complexity of operation. 
 
In Fig. 9 and 10 the distribution of the total mass for 
each level of radiological complexity of operation at 3 
cooling times is showed for both irradiation scenarios.  
 
 
 
Fig. 8. Distribution of the total mass (in tons) for each 
level of radiological complexity of operation under 
DEMO scenario at 3 cooling times (1, 10 and 100 y).  
FUSION SCIENCE AND TECHNOLOGY        VOL. 60        AUG. 2011 741
Catalán et al.        ACTIVATION ANALYSIS OF DCLL BLANKET FOR DEMO REACTOR
 
 
Fig. 9. Distribution of the total mass (in tons) for each 
level of radiological complexity of operations under plant 
scenario and 3 times of cooling time (1, 10 and 100 y). 
All produced wastes are in 1-3 levels after 100 years 
of cooling time in both irradiation scenarios, avoiding the 
troublesome level 4.  
III.C.2. Activity-level classification. 
Following the IAEA classification above described 
the table III shows the mass in tons of LILW and HLW 
for the materials of the blanket and the VV. It can be seen 
that beyond 10 years of cooling time all the radioactive 
wastes are in LILW. 
  
TABLE III. Mass in tons for both irradiation scenarios. 
Level 1 y 10 y 100 y 1000 y 
LILW (DEMO) 12256 12443 12443 12443 
HLW (DEMO) 187 0 0 0 
LILW (Plant) 61112 64335 64335 64335 
HLW (Plant) 3223 0 0 0 
IV. CONCLUSIONS 
 
Assessment of the radioactive waste generated in a 
conceptual design of a DCLL blanket has been performed 
taking into account its radiological complexity of 
operations and its activity-level classification.  
For both DEMO and power plant scenarios, all the 
radioactive wastes of the blanket and VV are LILW after 
one year of cooling time, and all the wastes are in 1-3 
complexity of operation levels after 100 years of cooling 
time, avoiding the troublesome level 4. None of the 
materials that compound the blanket fulfills the limits of 
recycling in foundries and unconditional clearance.  
The bioshielding is allowed for unconditional 
clearance before the first year of cooling for both 
irradiation scenarios and the cryostat can be recycled in 
foundries between 15-50 years for DEMO scenario and at 
100 years for the power plant scenario.  
Ongoing work on the study of this design with respect 
to waste management is focused on performing a refined 
analysis using a more realistic neutronic model and a 
thorough study about the recycling and disposal routes 
under a particular legislation like the Spanish one.  
 
ACKNOWLEDGMENTS 
 
This work is founded by the Spanish National Project 
on Breeding Blanket Technologies TECNO FUS through 
CONSOLIDER-INGENIO 2010 Programme. It has been 
also partially supported by Plan Nacional I + D + I (2008 
-2011) Fusion Nuclear ENE2008-06403-C06-02 MICINN 
(Spain).  
 
REFERENCES 
 
1. ³A conceptual study of commercial fusion power 
plants´, Final Report of the European Fusion PPCS 
2005. 
2. L. EL-*8(%$/<HWDO ³$Q ,QWHJUDWHG$SSURDFK WR
the Back-End of the Fusion Materials C\FOH´Fusion 
Science and Technology, 52, 2 , 109 (2009). 
3. J. SANZ et al ³ACAB Inventory code for nuclear 
DSSOLFDWLRQV 8VHU¶V PDQXDO 9 ´ NEA-1839 
(2008). 
4. D. LÓPEZ and - 3 &$7$/È1 ³0&$&$%
8VHUV PDQXDO 9 ´ UNED, TECF3IR-GRAL-
001, January 2010. 
5. J. LATKOWSKI ³7$575($' 9 8VHUV
PDQXDO´, Laurence Livermore National Laboratory, 
UCRL-MA-122003, August 1995. 
6. J. P. CATALÁN et al ³Neutronic analysis of a dual 
He/LiPb coolant breeding blanket for DEMO´ 
Fusion Engineering Design (2011). 
7. Y. WU ³CAD-based interface programs for fusion 
neutron transport simulation´ Fusion Engineering 
Design (2009). 
8. R. LINDAU et al ³Present development status of 
Eurofer and ODS-Eurofer for application in blanket 
FRQFHSWV´ Fusion Engineering and Design, 75-79, 
989 (2005)  
9. L. GIANCARLI et al ³R&D issues for SiCf/SiC 
composites structural material in fusion power 
reactor blankets´ Fusion Engineering and Design, 
48, 509 (2000). 
10. H. YAMADA et al ³The neutron irradiation effect on 
mechanical properties of HIP joint material´ Journal 
of Nuclear Materials, 335, 33 (2004). 
11. ³&DWHJRUL]LQJ 2SHUDWLRQDO 5DGLRDFWLYH :DVWHV´
IAEA-TECDOC-1538. 
742 FUSION SCIENCE AND TECHNOLOGY        VOL. 60        AUG. 2011
Catalán et al.        ACTIVATION ANALYSIS OF DCLL BLANKET FOR DEMO REACTOR


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Activation Analysis for a He/LiPb dual Coolant Blanket for DEMO Reactor

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