- Email: [email protected]
On the nuclear response of the helium-cooled lithium lead test blanket module in ITER
Fusion Engineering and Design 75–79 (2005) 725–730
On the nuclear response of the helium-cooled lithium lead test blanket module in ITER P. Chiovaro ∗ , P.A. Di Maio, G. Vella Dipartimento di Ingegneria Nucleare, Universit`a di Palermo, Viale delle Scienze, 90128 Palermo, Italy Available online 10 August 2005
Abstract The helium-cooled lithium lead (HCLL) concept has been recently selected as one of the two European reference designs foreseen for the breeding blanket of a demonstration fusion reactor. In particular, within the framework of the research and development activities on this blanket line, an HCLL test blanket module (TBM) has to be designed and manufactured to be implemented in ITER. At the Department of Nuclear Engineering (DIN) of the University of Palermo, a research campaign has been carried out to investigate the nuclear response of HCLL-TBM inside ITER by a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of HCLL-TBM has been set-up and inserted into an ITER 3D semi-heterogeneous one that simulates realistically the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source has been adopted for the calculations. The main features of the HCLL-TBM nuclear response have been determined, paying particular attention to the deposited power and the tritium production rate together with the spatial distribution of their volumetric densities. The radiation damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rate. © 2005 Elsevier B.V. All rights reserved. Keywords: HCLL-blanket; Test blanket module; Neutronics
1. Introduction The helium-cooled lithium lead (HCLL) blanket is one of the two European breeding blanket lines selected for a demonstration fusion reactor. It relies on the use of Pb–Li liquid eutectic alloy both as the tritium breeder ∗ Corresponding author. Tel.: +39 031 232240; fax: +39 091 232215. E-mail address: [email protected] (P. Chiovaro).
0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.06.294
and the neutron multiplier, and on reduced-activation martensitic steel as the structural material. Helium at a pressure of 8 MPa and at a temperature ranging from 300 ◦ C to 500 ◦ C is envisaged as the coolant [1]. Within the framework of the research and development activities focused on that blanket line, the design and manufacturing of proper test blanket modules (TBMs) to be implemented in ITER play a pivotal role.
726
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730
In particular, at the Department of Nuclear Engineering (DIN) of the University of Palermo, a research campaign has been started with the specific aim of investigating the nuclear response of HCLL-TBM when operating inside ITER, to provide useful data for the determination of both its thermal–hydraulic and thermal–mechanical behaviours.
2. Outline of HCLL-TBM HCLL-TBMs aim at representing a typical module of the HCLL blanket. They should be cased with a poloidal lay-out in an ITER equatorial port, being housed in a water-cooled steel frame directly supported by the vacuum vessel. In particular, Integral TBM (indicated in the following as HCLL-TBM) should integrate all relevant demo technologies and features and it should be tested during the high duty D–T plasma phase of ITER. HCLL-TBM has a poloidal height of 1.832 m, a toroidal width of 0.626 m and a radial depth of 0.585 m. From the structural point of view, it is characterized by a box-shaped structure composed of a segment box (SB) and a breeder zone (BZ) [1]. The segment box is a directly-cooled steel box having basically the function of Pb–Li container. It is mainly composed of a first wall (FW), two side walls (SWs), two caps and a set of back plates (BPs). SB is reinforced by 8 mm thick toroidal–radial and poloidal–radial stiffening plates (SPs) that subdivide breeder zone into 24 radial cells. They are arranged with pitches of 222 mm and 188 mm in poloidal and toroidal directions, respectively. The reference structural material is reduced-activation 9% Cr martensitic steel called EUROFER. A set of 20 cooling channels (14 mm × 15.6 mm) with a triple U-turn are grooved within FW and SWs with a poloidal pitch of 22.2 mm. Moreover, four double U-cooling channels (3 mm × 10 mm toroidal–radial; 3 mm × 8 mm poloidal–radial) are foreseen inside each SP. The breeder zone is occupied by the Pb–Li liquid eutectic alloy enriched up to 90% in Li6 . It has a modular structure articulated in 24 breeder units (BUs) housed in the aforementioned SB cells. Every BU has a poloidal height of 214 mm, a toroidal width of 180 mm and a radial depth of 348 mm. A set of five toroidal–radial cooling plates (CPs) subdivide each BU
into six channels in which Pb–Li alloy flows. Each CP is cooled by eight double U-channels (4 mm × 4.5 mm). Further details on HCLL-TBM lay-out can be found in ref. [1].
3. Nuclear analyses A detailed 3D study of the HCLL-TBM nuclear response in ITER has been performed by means of the Monte Carlo method, using the Monte Carlo N-Particle (MCNP) code version 4C and adopting the FENDL-2 transport cross section libraries [2,3]. In order to speed up calculations, the analyses have been carried out on a cluster of four workstations with heterogeneous operating systems by implementing the parallel virtual machine software. 3.1. The model A 3D semi-heterogeneous model of ITER has been adopted that realistically simulates a blanket sector of the whole reactor. In particular, it represents 1/18 (20◦ ) of the reactor toroidal extension, describing in detail the shielding blanket, the divertor cassette, the magnet system and the vacuum vessel with its three major ports and the cryostat. Two symmetry conditions have been imposed at the model toroidal boundaries to simulate reactor’s continuity in that direction. The Gaussianshaped neutron source developed by the Physics Unit of Naka ITER Joint Work Site [4] has been taken into account to simulate the D–T plasma of ITER. A detailed 3D heterogeneous model of HCLL-TBM has been set-up and it has been inserted into an appropriate homogeneous model of the steel supporting frame properly located inside an equatorial port of ITER outboard blanket. In particular, HCLL-TBM has been modelled in an almost fully heterogeneous way (Figs. 1 and 2) with some simplifications as far as SPs and CPs are concerned. Those have been modelled as proper homogeneous mixtures of EUROFER and helium, accordingly to their volume fractions within CPs and SPs. Pb–17Li has been assumed as the tritium breeder. Concerning the steel frame, it has been considered as a homogeneous mixture of water and AISI 316L (20% and 80% in weight, respectively).
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730
727
Table 1 Distribution of the nuclear power deposited within HCLL-TBM (kW) BZ Pb–17Li CPs SB Poloidal–radial SPs Toroidal–radial SPs FW SWs Top cap Bottom cap BPs Total
390.93 28.04 5.99 6.76 86.53 55.50 11.55 12.04 14.47 611.81
photonic deposited power, tritium production and radiation damage of the structural material. The analyses have been carried out by simulating a large number of histories (2 × 107 ) so that the results obtained are affected by statistical uncertainties lower than 1%.
Fig. 1. Poloidal–radial section of the HCLL-TBM model.
3.2. The results The HCLL-TBM nuclear response has been investigated paying particular attention to neutronic and
Fig. 2. Toroidal–radial section of the HCLL-TBM model.
3.2.1. Neutronic and photonic power deposition The power deposited in the module by neutrons and photons has been evaluated in order to provide useful data for the investigation of the HCLL-TBM thermal–hydraulic performances. It has been estimated that a total power of 611.8 kW is released within the module and a detailed description of its distribution is reported in Table 1. The spatial distribution of the volumetric density of the nuclear power deposited both in SB and BZ has been evaluated to allow studying HCLL-TBM thermal–mechanical performances. As the variations expected along the toroidal direction are negligible, only the radial and poloidal distributions have been determined. To that purpose, the model has been subdivided in 14 radial volumes and, as far as the Pb–17Li is concerned, each of them has been further subdivided in three poloidal sub-volumes. The upper and lower ones comprise two BUs while the central one the remaining four. The radial profiles obtained are shown in Figs. 3 and 4. As expected, the deposited power densities reach their highest values near the plasma-facing region both in SB and BZ, decreasing significantly along the radial direction. In particular, within FW a value of 4.9 W cm−3 has been calculated.
728
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730 Table 2 Best fitting coefficients
Fig. 3. Nuclear power density distribution in Pb–17Li eutectic alloy.
The radial distributions have been fitted accordingly to the least-square method throughout the following functional form: q (r) = α exp(−βr) + γ exp(−δr) where r represents the radial distance in centimetres from the FW plasma-facing surface. The minimum correlation factor obtained is higher than 0.998 and the best fitting functions determined, whether properly integrated, save the overall deposited power. The functions have been reported in Figs. 3 and 4 in comparison with the code predictions, while their coefficients are reported in Table 2. Finally, the average neutron wall loading along HCLL-TBM plasma facing surface has been estimated to be 0.74 MWm−2 . 3.2.2. Tritium production As one of the HCLL-TBM main goals is to test the tritium breeding, recovery and confinement capability of such a kind of blanket line, the daily tritium production together with the radial distribution of the volumetric density of the tritium production rate have been investigated.
Fig. 4. Nuclear power density distribution in the segment box.
Zone
α
β
γ
δ
Pb–17Li Upper q T
100.10693 172.38562
1.144311 1.3191614
5.1048387 2.4173216
0.09370264 0.05586956
Central q T
99.649142 172.17448
1.1396933 1.3212275
5.3836525 2.4425275
0.0890884 0.05001622
Lower q T
102.27753 152.70634
1.1606674 1.2832231
5.447592 2.4498479
0.09515334 0.05483909
1.2255282
0.039155105
5.046685
0.18108867
2.1929478
0.0547928
3.5665707
0.17268503
1.4121318
0.04365287
4.4913213
0.14090156
SWs
q
Caps Top q Bottom q
The daily tritium production depends obviously on the lithium enrichment of Pb–17Li and on the ITER duty cycle. Assuming a nominal Li6 enrichment (90%) and a duty cycle of 0.22, the daily tritium production has been calculated to be 13.13 mg day−1 . This value is lower than the one calculated with respect to the water-cooled lithium lead TBM [5], as a consequence of both the harder neutronic spectrum and the smaller BZ radial depth that characterize HCLL-TBM. In order to perform tritium permeation analyses, the radial distributions of the volumetric density of the tritium production rate have been determined in the upper, central and lower regions of BZ. The distributions obtained are shown in Fig. 5 together with their
Fig. 5. Distribution of the tritium production rate density in the segment box.
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730
729
best fitting functions (T ) that have the same analytical form of the power density ones and the coefficients of which are shown in Table 2. 3.2.3. Radiation damage During plant operation, the highly energetic fusion neutrons coming out from plasma continuously interact with the structural material atoms inducing 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 a result of various nuclear reactions mainly of (n, p) and (n, ␣) kind. While the hydrogen isotopes diffuse out of the metallic lattice or form metal hydrides, ␣-particles remain trapped in the metal and generate helium gas bubbles. These processes lead to unfavourable changes of mechanical properties (such as embrittlement), limit the lifetime of the structural material and affect its reweldability [5]. In order to investigate the level of the HCLL-TBM material damage due to the first mechanism, it has been evaluated the displacement per atom (DPA) within the SB by adopting the displacement cross sections for iron taken from ASTM standards [6]. The DPA distribution along the radial depth of the HCLL-TBM structural material (Fig. 6) have been evaluated by assuming a reactor duty cycle of 0.22 and by supposing a full power pulsed plant operation during a whole year. The maximum is obviously reached within FW and it is 1.12 DPA year−1 . In order to estimate the effect of the second damage mechanism, helium and hydrogen production rates have been also evaluated along the radial direction within the SB structural material. The profiles obtained
Fig. 6. DPA distribution in the segment box.
Fig. 7. Distribution of He and H production rate in the segment box.
are reported in Fig. 7, where it is possible to observe that maxima are reached of about 11.7 and 34.5 appm year−1 for helium and hydrogen, respectively.
4. Conclusions A detailed investigation of the HCLL-TBM nuclear response in ITER has been performed by means of 3D-Monte Carlo neutronic and photonic analyses. A 3D heterogeneous model of HCLL-TBM has been set-up taking into account EUROFER as the structural material. It has been inserted into an ITER semiheterogeneous model with a proper D–T plasma neutron source. The main features of the HCLL-TBM nuclear response have been determined, focusing the attention onto power deposition, tritium production and radiation-induced material damage. The nuclear power deposited in HCLL–TBM has been estimated to be about 612 kW and its detailed spatial distribution has been evaluated. A particular attention has been paid to the radial distributions of the volumetric density of the nuclear deposited power, finding out their relevant best fitting functions that could be useful for the study of the module thermal–mechanical performances. The daily production of tritium together with the radial distribution of its production rate have been evaluated, observing that the former reaches a value slightly higher than 13 mg day−1 . As far as the radiation damage is concerned, both 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 achieved in the FW proximity, where the neutron fluence is higher. Those maxima, in the hypothesis of 1 year of full pulsed power
730
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730
operation, have been estimated to be 1.12 DPA year−1 and about 11.7 and 34.5 appm year−1 for helium and hydrogen, respectively. The frame configuration (geometry and composition) could have a deep impact on the module nuclear response that has not been investigated in this study. That impact will be clarified in the further developments of the DIN research campaign.
[2] [3]
[4]
[5]
References [1] G. Rampal, A. Li Puma, Y. Poitevin, E. Rigal, J. Szczepanski, C. Boudot, HCLL-TBM for ITER design studies, in: Proceedings of
[6]
the 23rd Symposium on Fusion Technology, Venezia, September 20–24, 2004. J.F. Briesmeister, MCNP, A General Monte Carlo N-Particle Transport Code, LA-12625-M, Version 4C, 2000. M. Herman, A.B. Pashchenko (Eds.), Extension and improvement of the FENDL library for fusion applications (FENDL-2), Report on a Meeting Held in Vienna, Austria, March 3–7, 1997, Report INDC (NDS)-373, 1997. G. Ruvutuso, H. Ida, L. Petrizzi, Updated of basic 3D Model of ITER for Monte Carlo Nuclear Analyses with MCNP, ITER Joint Work Site, November 14, 2000. P. Chiovaro, P.A. Di Maio, E. Oliveri, G. Vella, On the nuclear response of the water-cooled Pb–17Li test blanket module for ITER-FEAT, Fusion Eng. Des. 69 (2003) 469–477. 1994 Annual Book of ASTM Standards, vol. 12.02 Nuclear (II) Solar and Geothermal Energy, E 693–694.
On the nuclear response of the helium-cooled lithium lead test blanket module in ITER P. Chiovaro ∗ , P.A. Di Maio, G. Vella Dipartimento di Ingegneria Nucleare, Universit`a di Palermo, Viale delle Scienze, 90128 Palermo, Italy Available online 10 August 2005
Abstract The helium-cooled lithium lead (HCLL) concept has been recently selected as one of the two European reference designs foreseen for the breeding blanket of a demonstration fusion reactor. In particular, within the framework of the research and development activities on this blanket line, an HCLL test blanket module (TBM) has to be designed and manufactured to be implemented in ITER. At the Department of Nuclear Engineering (DIN) of the University of Palermo, a research campaign has been carried out to investigate the nuclear response of HCLL-TBM inside ITER by a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of HCLL-TBM has been set-up and inserted into an ITER 3D semi-heterogeneous one that simulates realistically the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source has been adopted for the calculations. The main features of the HCLL-TBM nuclear response have been determined, paying particular attention to the deposited power and the tritium production rate together with the spatial distribution of their volumetric densities. The radiation damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rate. © 2005 Elsevier B.V. All rights reserved. Keywords: HCLL-blanket; Test blanket module; Neutronics
1. Introduction The helium-cooled lithium lead (HCLL) blanket is one of the two European breeding blanket lines selected for a demonstration fusion reactor. It relies on the use of Pb–Li liquid eutectic alloy both as the tritium breeder ∗ Corresponding author. Tel.: +39 031 232240; fax: +39 091 232215. E-mail address: [email protected] (P. Chiovaro).
0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.06.294
and the neutron multiplier, and on reduced-activation martensitic steel as the structural material. Helium at a pressure of 8 MPa and at a temperature ranging from 300 ◦ C to 500 ◦ C is envisaged as the coolant [1]. Within the framework of the research and development activities focused on that blanket line, the design and manufacturing of proper test blanket modules (TBMs) to be implemented in ITER play a pivotal role.
726
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730
In particular, at the Department of Nuclear Engineering (DIN) of the University of Palermo, a research campaign has been started with the specific aim of investigating the nuclear response of HCLL-TBM when operating inside ITER, to provide useful data for the determination of both its thermal–hydraulic and thermal–mechanical behaviours.
2. Outline of HCLL-TBM HCLL-TBMs aim at representing a typical module of the HCLL blanket. They should be cased with a poloidal lay-out in an ITER equatorial port, being housed in a water-cooled steel frame directly supported by the vacuum vessel. In particular, Integral TBM (indicated in the following as HCLL-TBM) should integrate all relevant demo technologies and features and it should be tested during the high duty D–T plasma phase of ITER. HCLL-TBM has a poloidal height of 1.832 m, a toroidal width of 0.626 m and a radial depth of 0.585 m. From the structural point of view, it is characterized by a box-shaped structure composed of a segment box (SB) and a breeder zone (BZ) [1]. The segment box is a directly-cooled steel box having basically the function of Pb–Li container. It is mainly composed of a first wall (FW), two side walls (SWs), two caps and a set of back plates (BPs). SB is reinforced by 8 mm thick toroidal–radial and poloidal–radial stiffening plates (SPs) that subdivide breeder zone into 24 radial cells. They are arranged with pitches of 222 mm and 188 mm in poloidal and toroidal directions, respectively. The reference structural material is reduced-activation 9% Cr martensitic steel called EUROFER. A set of 20 cooling channels (14 mm × 15.6 mm) with a triple U-turn are grooved within FW and SWs with a poloidal pitch of 22.2 mm. Moreover, four double U-cooling channels (3 mm × 10 mm toroidal–radial; 3 mm × 8 mm poloidal–radial) are foreseen inside each SP. The breeder zone is occupied by the Pb–Li liquid eutectic alloy enriched up to 90% in Li6 . It has a modular structure articulated in 24 breeder units (BUs) housed in the aforementioned SB cells. Every BU has a poloidal height of 214 mm, a toroidal width of 180 mm and a radial depth of 348 mm. A set of five toroidal–radial cooling plates (CPs) subdivide each BU
into six channels in which Pb–Li alloy flows. Each CP is cooled by eight double U-channels (4 mm × 4.5 mm). Further details on HCLL-TBM lay-out can be found in ref. [1].
3. Nuclear analyses A detailed 3D study of the HCLL-TBM nuclear response in ITER has been performed by means of the Monte Carlo method, using the Monte Carlo N-Particle (MCNP) code version 4C and adopting the FENDL-2 transport cross section libraries [2,3]. In order to speed up calculations, the analyses have been carried out on a cluster of four workstations with heterogeneous operating systems by implementing the parallel virtual machine software. 3.1. The model A 3D semi-heterogeneous model of ITER has been adopted that realistically simulates a blanket sector of the whole reactor. In particular, it represents 1/18 (20◦ ) of the reactor toroidal extension, describing in detail the shielding blanket, the divertor cassette, the magnet system and the vacuum vessel with its three major ports and the cryostat. Two symmetry conditions have been imposed at the model toroidal boundaries to simulate reactor’s continuity in that direction. The Gaussianshaped neutron source developed by the Physics Unit of Naka ITER Joint Work Site [4] has been taken into account to simulate the D–T plasma of ITER. A detailed 3D heterogeneous model of HCLL-TBM has been set-up and it has been inserted into an appropriate homogeneous model of the steel supporting frame properly located inside an equatorial port of ITER outboard blanket. In particular, HCLL-TBM has been modelled in an almost fully heterogeneous way (Figs. 1 and 2) with some simplifications as far as SPs and CPs are concerned. Those have been modelled as proper homogeneous mixtures of EUROFER and helium, accordingly to their volume fractions within CPs and SPs. Pb–17Li has been assumed as the tritium breeder. Concerning the steel frame, it has been considered as a homogeneous mixture of water and AISI 316L (20% and 80% in weight, respectively).
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730
727
Table 1 Distribution of the nuclear power deposited within HCLL-TBM (kW) BZ Pb–17Li CPs SB Poloidal–radial SPs Toroidal–radial SPs FW SWs Top cap Bottom cap BPs Total
390.93 28.04 5.99 6.76 86.53 55.50 11.55 12.04 14.47 611.81
photonic deposited power, tritium production and radiation damage of the structural material. The analyses have been carried out by simulating a large number of histories (2 × 107 ) so that the results obtained are affected by statistical uncertainties lower than 1%.
Fig. 1. Poloidal–radial section of the HCLL-TBM model.
3.2. The results The HCLL-TBM nuclear response has been investigated paying particular attention to neutronic and
Fig. 2. Toroidal–radial section of the HCLL-TBM model.
3.2.1. Neutronic and photonic power deposition The power deposited in the module by neutrons and photons has been evaluated in order to provide useful data for the investigation of the HCLL-TBM thermal–hydraulic performances. It has been estimated that a total power of 611.8 kW is released within the module and a detailed description of its distribution is reported in Table 1. The spatial distribution of the volumetric density of the nuclear power deposited both in SB and BZ has been evaluated to allow studying HCLL-TBM thermal–mechanical performances. As the variations expected along the toroidal direction are negligible, only the radial and poloidal distributions have been determined. To that purpose, the model has been subdivided in 14 radial volumes and, as far as the Pb–17Li is concerned, each of them has been further subdivided in three poloidal sub-volumes. The upper and lower ones comprise two BUs while the central one the remaining four. The radial profiles obtained are shown in Figs. 3 and 4. As expected, the deposited power densities reach their highest values near the plasma-facing region both in SB and BZ, decreasing significantly along the radial direction. In particular, within FW a value of 4.9 W cm−3 has been calculated.
728
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730 Table 2 Best fitting coefficients
Fig. 3. Nuclear power density distribution in Pb–17Li eutectic alloy.
The radial distributions have been fitted accordingly to the least-square method throughout the following functional form: q (r) = α exp(−βr) + γ exp(−δr) where r represents the radial distance in centimetres from the FW plasma-facing surface. The minimum correlation factor obtained is higher than 0.998 and the best fitting functions determined, whether properly integrated, save the overall deposited power. The functions have been reported in Figs. 3 and 4 in comparison with the code predictions, while their coefficients are reported in Table 2. Finally, the average neutron wall loading along HCLL-TBM plasma facing surface has been estimated to be 0.74 MWm−2 . 3.2.2. Tritium production As one of the HCLL-TBM main goals is to test the tritium breeding, recovery and confinement capability of such a kind of blanket line, the daily tritium production together with the radial distribution of the volumetric density of the tritium production rate have been investigated.
Fig. 4. Nuclear power density distribution in the segment box.
Zone
α
β
γ
δ
Pb–17Li Upper q T
100.10693 172.38562
1.144311 1.3191614
5.1048387 2.4173216
0.09370264 0.05586956
Central q T
99.649142 172.17448
1.1396933 1.3212275
5.3836525 2.4425275
0.0890884 0.05001622
Lower q T
102.27753 152.70634
1.1606674 1.2832231
5.447592 2.4498479
0.09515334 0.05483909
1.2255282
0.039155105
5.046685
0.18108867
2.1929478
0.0547928
3.5665707
0.17268503
1.4121318
0.04365287
4.4913213
0.14090156
SWs
q
Caps Top q Bottom q
The daily tritium production depends obviously on the lithium enrichment of Pb–17Li and on the ITER duty cycle. Assuming a nominal Li6 enrichment (90%) and a duty cycle of 0.22, the daily tritium production has been calculated to be 13.13 mg day−1 . This value is lower than the one calculated with respect to the water-cooled lithium lead TBM [5], as a consequence of both the harder neutronic spectrum and the smaller BZ radial depth that characterize HCLL-TBM. In order to perform tritium permeation analyses, the radial distributions of the volumetric density of the tritium production rate have been determined in the upper, central and lower regions of BZ. The distributions obtained are shown in Fig. 5 together with their
Fig. 5. Distribution of the tritium production rate density in the segment box.
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730
729
best fitting functions (T ) that have the same analytical form of the power density ones and the coefficients of which are shown in Table 2. 3.2.3. Radiation damage During plant operation, the highly energetic fusion neutrons coming out from plasma continuously interact with the structural material atoms inducing 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 a result of various nuclear reactions mainly of (n, p) and (n, ␣) kind. While the hydrogen isotopes diffuse out of the metallic lattice or form metal hydrides, ␣-particles remain trapped in the metal and generate helium gas bubbles. These processes lead to unfavourable changes of mechanical properties (such as embrittlement), limit the lifetime of the structural material and affect its reweldability [5]. In order to investigate the level of the HCLL-TBM material damage due to the first mechanism, it has been evaluated the displacement per atom (DPA) within the SB by adopting the displacement cross sections for iron taken from ASTM standards [6]. The DPA distribution along the radial depth of the HCLL-TBM structural material (Fig. 6) have been evaluated by assuming a reactor duty cycle of 0.22 and by supposing a full power pulsed plant operation during a whole year. The maximum is obviously reached within FW and it is 1.12 DPA year−1 . In order to estimate the effect of the second damage mechanism, helium and hydrogen production rates have been also evaluated along the radial direction within the SB structural material. The profiles obtained
Fig. 6. DPA distribution in the segment box.
Fig. 7. Distribution of He and H production rate in the segment box.
are reported in Fig. 7, where it is possible to observe that maxima are reached of about 11.7 and 34.5 appm year−1 for helium and hydrogen, respectively.
4. Conclusions A detailed investigation of the HCLL-TBM nuclear response in ITER has been performed by means of 3D-Monte Carlo neutronic and photonic analyses. A 3D heterogeneous model of HCLL-TBM has been set-up taking into account EUROFER as the structural material. It has been inserted into an ITER semiheterogeneous model with a proper D–T plasma neutron source. The main features of the HCLL-TBM nuclear response have been determined, focusing the attention onto power deposition, tritium production and radiation-induced material damage. The nuclear power deposited in HCLL–TBM has been estimated to be about 612 kW and its detailed spatial distribution has been evaluated. A particular attention has been paid to the radial distributions of the volumetric density of the nuclear deposited power, finding out their relevant best fitting functions that could be useful for the study of the module thermal–mechanical performances. The daily production of tritium together with the radial distribution of its production rate have been evaluated, observing that the former reaches a value slightly higher than 13 mg day−1 . As far as the radiation damage is concerned, both 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 achieved in the FW proximity, where the neutron fluence is higher. Those maxima, in the hypothesis of 1 year of full pulsed power
730
P. Chiovaro et al. / Fusion Engineering and Design 75–79 (2005) 725–730
operation, have been estimated to be 1.12 DPA year−1 and about 11.7 and 34.5 appm year−1 for helium and hydrogen, respectively. The frame configuration (geometry and composition) could have a deep impact on the module nuclear response that has not been investigated in this study. That impact will be clarified in the further developments of the DIN research campaign.
[2] [3]
[4]
[5]
References [1] G. Rampal, A. Li Puma, Y. Poitevin, E. Rigal, J. Szczepanski, C. Boudot, HCLL-TBM for ITER design studies, in: Proceedings of
[6]
the 23rd Symposium on Fusion Technology, Venezia, September 20–24, 2004. J.F. Briesmeister, MCNP, A General Monte Carlo N-Particle Transport Code, LA-12625-M, Version 4C, 2000. M. Herman, A.B. Pashchenko (Eds.), Extension and improvement of the FENDL library for fusion applications (FENDL-2), Report on a Meeting Held in Vienna, Austria, March 3–7, 1997, Report INDC (NDS)-373, 1997. G. Ruvutuso, H. Ida, L. Petrizzi, Updated of basic 3D Model of ITER for Monte Carlo Nuclear Analyses with MCNP, ITER Joint Work Site, November 14, 2000. P. Chiovaro, P.A. Di Maio, E. Oliveri, G. Vella, On the nuclear response of the water-cooled Pb–17Li test blanket module for ITER-FEAT, Fusion Eng. Des. 69 (2003) 469–477. 1994 Annual Book of ASTM Standards, vol. 12.02 Nuclear (II) Solar and Geothermal Energy, E 693–694.