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Neutronics studies for the design of the European DEMO vacuum vessel
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ARTICLE IN PRESS Fusion Engineering and Design xxx (2016) xxx–xxx
Contents lists available at ScienceDirect
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Neutronics studies for the design of the European DEMO vacuum vessel Davide Flammini a,∗ , Rosaria Villari a , Fabio Moro a , Aldo Pizzuto a , Christian Bachmann b a b
ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome, Italy EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching, Germany
h i g h l i g h t s • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial • • • •
plane. Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel.
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Article history: Received 24 July 2015 Received in revised form 26 January 2016 Accepted 28 January 2016 Available online xxx Keywords: Vacuum vessel DEMO Nuclear heating Damage Neutronics
a b s t r a c t The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm3 ), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided. © 2016 Davide Flammini. Published by Elsevier B.V. All rights reserved.
1. Introduction The conceptual design activity of the Demonstration Fusion Power Reactor (DEMO) has recently been launched in the Power Plant Physics and Technology (PPPT) programme within the EUROfusion Consortium. Neutronics studies, fundamental for the nuclear design of DEMO and its components, are in progress. This paper is devoted to the neutronics calculations performed for the design of the vacuum vessel.
The nuclear heat load on the inner shell of the vacuum vessel (VV) will be higher behind the poloidal gaps between adjacent blanket segments with respect to the zone behind the blanket. Thermal loads are caused by temperature gradients inside the VV structure and the nuclear heating gradient might be critical [1]. Furthermore the neutron streaming through the gaps increases the local damage and He-production, thus can have impact on the VV lifetime and re-weldability. The aim of this work is to define a preliminary heat load specification on the VV in the inboard equatorial plane for different sizes of the poloidal gap between adjacent blanket segments.
∗ Corresponding author. E-mail addresses: davide.fl[email protected], davidefl[email protected] (D. Flammini). http://dx.doi.org/10.1016/j.fusengdes.2016.01.075 0920-3796/© 2016 Davide Flammini. Published by Elsevier B.V. All rights reserved.
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075
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2 Table 1 Main DEMO parameters defined in 2103. Major radius (m)
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Minor radius (m) Plasma elongation Plasma triangularity Plasma peaking factor Fusion power (MW)
2.25 1.66 0.33 1.3 1572
Fig. 2. MCNP model of the VV designed in order to reproduce the CAD model shown in figure 1 for DEMO HCLL. In the inset is also shown DEMO HCPB model with the homogenized blanket.
Fig. 1. CAD model of the inboard VV.
2. Methodology The nuclear quantities of interest have been calculated for the Helium Cooled Lithium Lead (HCLL) and HCPB DEMO model with MCNP5 v1.6 [2], using FENDL 2.1 nuclear data library [3]. The model used in this work is the 3-D MCNP DEMO model developed by KIT and updated by ENEA with a detailed description of the HCLL blanket modules in 2013[4]. It represents an 11.25◦ sector including VV, ports, divertor, coils and a banana-shaped void filled with the HCLL blanket modules and manifolds. The neutron plasma source was simulated making use of the specially developed source subroutine [5], describing a plasma scenario with the parameters reported in Table 1. With respect to the DEMO model of 2013 the fusion power has been set to 1572 MW. The DEMO model of 2013 has been updated for this study according to the recent design of the inboard VV shown in Fig. 1. The structural part of the VV is composed of two 60 mm thick cylindrical walls that enclose the inner part composed of 40 mm thick ribs [6,7]. The total radial thickness is 597 mm. The ribs and the walls are made of SS-316, additional plates or shields are not considered in the CAD. The MCNP model of the VV has been modified in order to reproduce the radial build-up of the CAD model of the VV. In Fig. 2 the new MCNP model of the VV is shown. In this preliminary study the ribs have not been modeled and an internal filling mixture of 80% SS316 steel and 20% water was assumed. A poloidal gap has then been inserted in the central part of the blanket module, representing the gap between adjacent segments. Four gap configurations have been analyzed: 1, 2, 3 and 5 cm wide gaps. The gaps in the poloidal direction are centered on the equatorial plane (see Fig. 3). A calculation without the gap has been performed as well. Moreover, in order to assess the differences between the blanket concepts, additional calculations were also performed with the Helium Cooled Pebble Bed (HCPB) blanket concept with 2 cm and 5 cm wide poloidal gaps. The HCPB DEMO model (with homogeneous breeder mixture) [8] has been developed by KIT and modified at the VV inboard sector.
Fig. 3. Poloidal section of the MCNP model, showing the gap in the poloidal direction and the sampling zone used in the calculations.
3. Calculations description Spatial distributions of nuclear heating (NH), neutron damage (in DPA) and helium production (in appm) in steel have been calculated, as well as neutron and photon fluxes. The FMESH tally feature of MCNP has been applied to evaluate the nuclear responses using a cylindrical mesh superimposed to the geometry. Data are averaged over 50 cm in the poloidal region close to the equatorial plane. This sampling zone is shown in Fig. 3. Finest angular binning is used in the region close to the gap. The radial bins have been chosen to exactly match the surfaces (cylinders) describing the VV. The mesh is shown in Fig. 4. Variance reduction techniques (weight windows) have been used to obtain results with reasonably low statistical uncertainties in the whole VV inboard region. Separate calculations have been performed for the different gaps size (from 0 to 5 cm) and blanket modules (HCPB and HCLL). NH and fluxes
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075
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Fig. 4. Cylindrical mesh used for the calculations.
are normalized to a total neutron yield of 5.567 × 1020 n/s, corresponding to 1572 MW of fusion power and a neutron wall load of 0.99 MW/m2 . DPA and He production have been normalized to 1 Full Power Year (FPY) operation. Although the transport of the particles have been performed considering the real composition of all the elements of the geometry, the calculations of NH, DPA and He production refer to SS-316 components. 4. Results The nuclear heating in SS-316 components of the VV has been calculated as the sum of the nuclear heating due to neutrons and secondary gammas. In Fig. 5 the NH in the inner shell and first layer behind inner shell are reported as a function of (toroidal angle). The NH results to be larger in the first VV layer with respect to the VV inner shell. This effect is due to the steel–water interface that causes an increase in the gamma generation [9], whose contribution dominates the NH on steel. The overall maximum value of the NH is obtained for the 5 cm gap at 10 cm penetration depth (first VV layer behind the VV inner shell) at the center of the gap and it is 1.32 W/cm3 . The maximum values for the 2 and 3 cm gaps are still found at the center of the gap and in the VV first layer: these result 1.05 and 1.13 W/cm3 for the 2 and 3 cm gaps, respectively. For the configuration with 1 cm gap the peak is no more observed. The NH is quite constant as a function of the toroidal angle and the average value is 0.99 W/cm3 . The NH is still quite large even for configuration without the gap: the average value is 0.97 W/cm3 . It can be noted that reducing the gap the NH on the side is higher than in the middle: this is a geometrical effect due to the fact that the cylindrical shape of VV is not reproduced in the blanket segmentation. Moreover, by comparing the NH values at far from the gap for different gap sizes, it can be noted that the gap has a small but not negligible contribution even far from the gap (the bin at = 10.7◦ is about 50 cm far from the central one): it is ≈10% larger with 5 cm gap with respect to the case without gap. Considering the averaged values over of the NH, it decreases up to 3 orders of magnitude, as a function of the penetration depth inside the VV for all the configurations. The peak value in the outer shell of the VV for the 5 and 3 cm wide gap configurations is approximately 9 × 10−4 W/cm3 and 7 × 10−4 W/cm3 , respectively. The outer shell is 4 cm far from the case of the toroidal field (TF) coil where there is a limit on the NH of 5 × 10−4 W/cm3 [10]. For these
Fig. 5. Results for the nuclear heating calculated in the VV inner shell (upper panel) and in the first layer of the internal part of the VV (lower panel).
Fig. 6. Calculated He production in SS- 316 (appm/FPY) at the VV inner shell of the VV for different gap configurations.
two configurations the NH is likely to be above the limit in the TF coil. In order to assess the impact of poloidal gap on the VV heating, the peaking factor has been calculated as the ratio of the NH behind the gap center and the value calculated far from the gap (the angular bin at 10.7◦ ). The peaking factor results to be maximum at the VV inner shell. For 5 cm gap it is 1.4 and it drops to 1.1 only for 2 cm gap. In the first layer of VV behind inner shell the effect is lower. Fig. 6 shows the results of the Helium production in the VV. The limit of the He production for re-weldability is 1 appm [10]. The
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075
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Fig. 7. Calculated damage in DPA/FPY in SS–316 at the inner shell of the VV for different gap configurations.
value of the He production far from the gap ranges between 0.11 and 0.12 appm/FPY. As a consequence, the limit for re-weldability would be reached in 9 FPY. However, major concern arises at position facing the gap center, where the calculated values of the helium production result to be 2.59, 1.54, 1.02 and 0.51 appm/FPY for 5, 3, 2 and 1 cm wide gap, respectively. As a result, the He production is largely influenced by the presence of the gap. This is because it is mostly generated by fast neutrons (E > 0.1 MeV) in steel [11]. It is worth nothing that in the present model there is no Boron content in steel. Boron is very important for He production because it has has a very high impact on it. To estimate the effect of the presence of Boron, assuming 10 ppm (0.001% weight, as recommended for the steel in ITER vacuum vessel cooling tubes), He production behind the blanket has to be multiplied by a factor of 2 [12]. DPA calculation in SS-316 has been performed as well in order to estimate the VV lifetime. In Fig. 7 the DPA at the inner shell of the VV is reported for all gap configurations. The maximum value of the DPA, obtained for the 5 cm gap in front of the gap center is 0.78, with a peaking factor of 1.8. Far from the gap the damage is < 0.5 dpa/FPY. Instead, it results 0.51 and 0.43 dpa/FPY for 2 and 1 cm gap configurations, respectively. Considering the limit of 2.75 dpa on VV steel [10], the 5 cm gap configuration does not meet the requirement, considering 6 FPY of operation. Instead with 1 and 2 cm wide gap the damage limit would be reached in 5.4 and 6.4 years, for 2 and 1 cm wide gaps, respectively. Same calculations were performed using HCPB DEMO homogeneous model for 2 and 5 cm wide gaps. Fig. 8 shows the comparison of the radial neutron flux between the HCLL and HCPB blanket for the case with the 2 cm wide gap, as a function of the radial distance from the first wall (FW). Differences arise even at the FW between the 2 different blankets: the total neutron fluxes result to be 7.3 × 1014 and 4.9 × 1014 n cm−2 s−1 , the fast neutron fluxes result to be 5.5 × 1014 and 3.3 × 1014 n cm−2 s−1 for the HCLL and for the HCPB, respectively. These results are consistent with that reported in [8] for the HCPB, taking into account for the proper normalization. Moreover, the HCPB blanket and manifolds provide a better shielding: the total neutron flux is attenuated by a factor 12.5 behind the blanket and manifold with respect to the flux at the FW with HCPB, compared to an attenuation of 5.4 with HCLL. In particular, for the fast neutron fluxes the attenuation factors are ≈22 with the HCPB blanket and ≈10 for the HCLL one. It should be noted, however, that the blanket model used for HCLL is heterogeneous, whereas the HCPB one is homogeneous and the heterogeneity effect could be not trivial.
Fig. 8. Neutron flux radial profile for HCPB and HCLL DEMO with 2 cm wide poloidal gap. Fast neutron (E > 0.1 MeV) and total neutrons fluxes are shown for both configurations.
Fig. 9. Nuclear heating of the VV as a function of the radial distance from its first wall. Comparison between the HCLL and HCPB blankets with a poloidal gap 2 cm wide.
These results on the neutron fluxes have direct consequences on the nuclear heating of the VV, leading to an overall smaller NH with the HCPB blanket, but with higher peaking factors. In Fig. 9 the mean value of the NH for the 2 cm gap has been shown for both blankets concept. The maximum values is still found for at 10 cm of penetration depth resulting in 0.30 W/cm3 for HCPB, compared with 1.02 W/cm3 for HCLL. The peaking factors are instead larger for HCPB (1.3) with respect to the HCLL. The 5 cm wide gap case have the same trend as the 2 cm wide gap one. The HCPB has a lower NH with respect to the HCLL (both for peak and mean values, 0.36 W/cm3 and 0.33 W/cm3 , respectively) with larger peaking factor (2.2) at the VV first layer. The outcome of this analysis is that the NH of the VV varies significantly for different blanket concepts. Comparing only HCLL and HCPB 2013 blanket configurations, differences on nuclear heating greater than a factor 3 are observed. Also, the peaking factors vary within 35% between the two examined configurations, showing that the peaking factor is strongly dependent on blanket shielding capability. In general it increases as the shielding capability is improved. In ITER, due to better shielding in front of VV, the peaking factor due to the poloidal gap is 3.3 [13]. Moreover, a more complete assessment has also to consider updated blanket concepts. 5. Conclusions The NH nuclear loads of the VV in DEMO HCLL blanket concept has been calculated for different poloidal gap sizes (0, 1, 2, 3 and
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075
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5 cm width) between blanket segments at the inboard equatorial region. The NH has been determined in toroidal and in radial direction. In poloidal direction the effect of the gap has been pointed out, showing a clear peak of the NH corresponding to the center of the gap. The peaking factors on NH in HCLL DEMO are small (within 1.4 with 5 cm gap width): this is because the gamma component in nuclear heating in steel is weakly affected by gap size and because the shielding capability of the 2013HCLL blanket is poor and high nuclear heating is obtained even in zones far from the gap and even without any gap. This study shows that with the HCLL blanket the effect of the poloidal gap could be important even for the TF coil where the limit of 5 × 10−4 W/cm3 could be exceeded with 3 and 5 cm wide gaps. The calculations with HCPB blanket show that the NH of the VV is lower by approximately a factor 3 with respect the HCLL blanket with the same gap configuration. This is due to a better shielding capability of HCPB with respect to HCLL. On the other hand the peaking factors obtained with HCPB are larger. The damage has also been calculated in the VV steel. The maximum value obtained for the 5 cm gap in front of the gap center is 0.78 dpa/FPY, with a peaking factor of 1.8 (5 cm gap). Considering the limit of 2.75 dpa, the only configuration that would not exceed the limit in the DEMO lifetime (∼6 fpy) is the 1 cm wide gap configuration. It may be worthwhile pointing out that for an updated design of the blanket the dpa result to be lower (<0.2 dpa/fpy) [14]. The helium production has been calculated as well. It results to have the largest peaking factors among the calculated quantity, being very sensitive to the gap size. The peak value for the 5 cm wide gap results 2.59 appm per FPY, exceeding the limit of 1 appm for reweldability in 0.39 FPY. Considering a gap size of 2 cm the Heproduction peak is 1.02 appm/FPY and the limit for reweldability is exceeded in ∼1 FPY. However, far from the gap, the he-production is much lower (0.12 appm/FPY) and no problem is foreseen in these zones. It could be noted that the present steel does not content boron which has an high impact on He-production.
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The computing resources and the related technical support used for this work have been provided by CRESCO/ENEAGRID High Performance Computing infrastructure and its staff [15]. CRESCO/ENEAGRID High Performance Computing infrastructure is funded by ENEA, see http://www.cresco.enea.it/english for information. References [1] G. Ramogida, et al., Thermal Structural Analysis Report of Vessel Inner Shell, EUROfusion report, 2014, EFDA D 2D4ZE9. [2] Mcnp, A general Monte Carlo N-particle Trans-port Code, Version 5, Los Alamos National Laboratory, 2013, https://mcnp.lanl.gov/ LA-UR-03-1987. [3] D. Lopez, Aldama, A. Trkov, FENDL-2.1, Update of an Evaluated Nuclear Data Library for Fusion Applications, International Atomic Energy Agency, Vienna, Austria, 2015, 18-July. [4] R. Villari, D. Flammini, F. Moro, Tritium Breeding Ratio Assessment for HCLL DEMO, EFDA D 2ME6F2, 30 May 2014 (https://idm.euro-fusion.org/ ?uid=2ME6F2&action=get document). [5] C. Fausser, et al., Tokamak D-T neutron source models for different plasma physics confinement modes, Fusion Eng. Des. 87 (2012) 787–792. [6] B. Meszaros, Methodology to Derive the Basic DEMO Tokamak ConfigurationCAD Geometry From System Code Studies, Fusion Eng. Design. (2014) (SOFT 2014). [7] C. Bachmann et al., Issues and strategies for DEMO in-vessel component integration, this Volume. [8] P. Pereslavtsev, L. Lu, U. Fischer, O. Bitz, Neutronic analyses of the HCPB DEMO reactor using a consistent integral approach, Fusion Eng. Des. 89 (2014) 1979–1983. [9] T.D. Bohm, M.E. Sawan, B. Smith, P.P.H. Wilson, Peaking in SS Nuclear Parameter at a Water Interface, UWFDM-1376, 2009. [10] U. Fischer, C. Bachmann, I. Palermo, P. Pereslavtsev, R. Villari, Neutronics requirements for a DEMO fusion power plant, in press, Fus. Eng. Des. (2015), http://dx.doi.org/10.1016/j.fusengdes.2015.02.029. [11] M. Shimada, M. Loughlin, M. Shute, Heat and Nuclear Load Specifications for ITER, IDM Number: ITER D 2LULDH v2.3, 16 November 2009, (https://user. iter.org/?uid=2luldh&action=get document). [12] R. Villari, WP12-DTM04-T06: Neutron transport calculation through the DEMO1 blanket (2LNDJ6 v1.1) (February 2013). [13] Heat and Nuclear Load Specifications for ITER, ITER D 2LULDH v2.3 (2009). [14] U. Fischer, et al., Neutronic Performance Issues of Breeder Blanket Options, ISFNT, 2015. [15] G. Ponti, et al., The role of medium size facilities in the HPC ecosystem: the case of the new CRESCO4 cluster integrated in the ENEAGRID infrastructure, Proceedings of the 2014 International Conference on High Performance Computing and Simulation, HPCS (2014) 1030–1033, 6903807.
Acknowledgments This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075
ARTICLE IN PRESS Fusion Engineering and Design xxx (2016) xxx–xxx
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Neutronics studies for the design of the European DEMO vacuum vessel Davide Flammini a,∗ , Rosaria Villari a , Fabio Moro a , Aldo Pizzuto a , Christian Bachmann b a b
ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome, Italy EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching, Germany
h i g h l i g h t s • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial • • • •
plane. Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel.
a r t i c l e
i n f o
Article history: Received 24 July 2015 Received in revised form 26 January 2016 Accepted 28 January 2016 Available online xxx Keywords: Vacuum vessel DEMO Nuclear heating Damage Neutronics
a b s t r a c t The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm3 ), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided. © 2016 Davide Flammini. Published by Elsevier B.V. All rights reserved.
1. Introduction The conceptual design activity of the Demonstration Fusion Power Reactor (DEMO) has recently been launched in the Power Plant Physics and Technology (PPPT) programme within the EUROfusion Consortium. Neutronics studies, fundamental for the nuclear design of DEMO and its components, are in progress. This paper is devoted to the neutronics calculations performed for the design of the vacuum vessel.
The nuclear heat load on the inner shell of the vacuum vessel (VV) will be higher behind the poloidal gaps between adjacent blanket segments with respect to the zone behind the blanket. Thermal loads are caused by temperature gradients inside the VV structure and the nuclear heating gradient might be critical [1]. Furthermore the neutron streaming through the gaps increases the local damage and He-production, thus can have impact on the VV lifetime and re-weldability. The aim of this work is to define a preliminary heat load specification on the VV in the inboard equatorial plane for different sizes of the poloidal gap between adjacent blanket segments.
∗ Corresponding author. E-mail addresses: davide.fl[email protected], davidefl[email protected] (D. Flammini). http://dx.doi.org/10.1016/j.fusengdes.2016.01.075 0920-3796/© 2016 Davide Flammini. Published by Elsevier B.V. All rights reserved.
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075
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ARTICLE IN PRESS
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D. Flammini et al. / Fusion Engineering and Design xxx (2016) xxx–xxx
2 Table 1 Main DEMO parameters defined in 2103. Major radius (m)
9
Minor radius (m) Plasma elongation Plasma triangularity Plasma peaking factor Fusion power (MW)
2.25 1.66 0.33 1.3 1572
Fig. 2. MCNP model of the VV designed in order to reproduce the CAD model shown in figure 1 for DEMO HCLL. In the inset is also shown DEMO HCPB model with the homogenized blanket.
Fig. 1. CAD model of the inboard VV.
2. Methodology The nuclear quantities of interest have been calculated for the Helium Cooled Lithium Lead (HCLL) and HCPB DEMO model with MCNP5 v1.6 [2], using FENDL 2.1 nuclear data library [3]. The model used in this work is the 3-D MCNP DEMO model developed by KIT and updated by ENEA with a detailed description of the HCLL blanket modules in 2013[4]. It represents an 11.25◦ sector including VV, ports, divertor, coils and a banana-shaped void filled with the HCLL blanket modules and manifolds. The neutron plasma source was simulated making use of the specially developed source subroutine [5], describing a plasma scenario with the parameters reported in Table 1. With respect to the DEMO model of 2013 the fusion power has been set to 1572 MW. The DEMO model of 2013 has been updated for this study according to the recent design of the inboard VV shown in Fig. 1. The structural part of the VV is composed of two 60 mm thick cylindrical walls that enclose the inner part composed of 40 mm thick ribs [6,7]. The total radial thickness is 597 mm. The ribs and the walls are made of SS-316, additional plates or shields are not considered in the CAD. The MCNP model of the VV has been modified in order to reproduce the radial build-up of the CAD model of the VV. In Fig. 2 the new MCNP model of the VV is shown. In this preliminary study the ribs have not been modeled and an internal filling mixture of 80% SS316 steel and 20% water was assumed. A poloidal gap has then been inserted in the central part of the blanket module, representing the gap between adjacent segments. Four gap configurations have been analyzed: 1, 2, 3 and 5 cm wide gaps. The gaps in the poloidal direction are centered on the equatorial plane (see Fig. 3). A calculation without the gap has been performed as well. Moreover, in order to assess the differences between the blanket concepts, additional calculations were also performed with the Helium Cooled Pebble Bed (HCPB) blanket concept with 2 cm and 5 cm wide poloidal gaps. The HCPB DEMO model (with homogeneous breeder mixture) [8] has been developed by KIT and modified at the VV inboard sector.
Fig. 3. Poloidal section of the MCNP model, showing the gap in the poloidal direction and the sampling zone used in the calculations.
3. Calculations description Spatial distributions of nuclear heating (NH), neutron damage (in DPA) and helium production (in appm) in steel have been calculated, as well as neutron and photon fluxes. The FMESH tally feature of MCNP has been applied to evaluate the nuclear responses using a cylindrical mesh superimposed to the geometry. Data are averaged over 50 cm in the poloidal region close to the equatorial plane. This sampling zone is shown in Fig. 3. Finest angular binning is used in the region close to the gap. The radial bins have been chosen to exactly match the surfaces (cylinders) describing the VV. The mesh is shown in Fig. 4. Variance reduction techniques (weight windows) have been used to obtain results with reasonably low statistical uncertainties in the whole VV inboard region. Separate calculations have been performed for the different gaps size (from 0 to 5 cm) and blanket modules (HCPB and HCLL). NH and fluxes
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075
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Fig. 4. Cylindrical mesh used for the calculations.
are normalized to a total neutron yield of 5.567 × 1020 n/s, corresponding to 1572 MW of fusion power and a neutron wall load of 0.99 MW/m2 . DPA and He production have been normalized to 1 Full Power Year (FPY) operation. Although the transport of the particles have been performed considering the real composition of all the elements of the geometry, the calculations of NH, DPA and He production refer to SS-316 components. 4. Results The nuclear heating in SS-316 components of the VV has been calculated as the sum of the nuclear heating due to neutrons and secondary gammas. In Fig. 5 the NH in the inner shell and first layer behind inner shell are reported as a function of (toroidal angle). The NH results to be larger in the first VV layer with respect to the VV inner shell. This effect is due to the steel–water interface that causes an increase in the gamma generation [9], whose contribution dominates the NH on steel. The overall maximum value of the NH is obtained for the 5 cm gap at 10 cm penetration depth (first VV layer behind the VV inner shell) at the center of the gap and it is 1.32 W/cm3 . The maximum values for the 2 and 3 cm gaps are still found at the center of the gap and in the VV first layer: these result 1.05 and 1.13 W/cm3 for the 2 and 3 cm gaps, respectively. For the configuration with 1 cm gap the peak is no more observed. The NH is quite constant as a function of the toroidal angle and the average value is 0.99 W/cm3 . The NH is still quite large even for configuration without the gap: the average value is 0.97 W/cm3 . It can be noted that reducing the gap the NH on the side is higher than in the middle: this is a geometrical effect due to the fact that the cylindrical shape of VV is not reproduced in the blanket segmentation. Moreover, by comparing the NH values at far from the gap for different gap sizes, it can be noted that the gap has a small but not negligible contribution even far from the gap (the bin at = 10.7◦ is about 50 cm far from the central one): it is ≈10% larger with 5 cm gap with respect to the case without gap. Considering the averaged values over of the NH, it decreases up to 3 orders of magnitude, as a function of the penetration depth inside the VV for all the configurations. The peak value in the outer shell of the VV for the 5 and 3 cm wide gap configurations is approximately 9 × 10−4 W/cm3 and 7 × 10−4 W/cm3 , respectively. The outer shell is 4 cm far from the case of the toroidal field (TF) coil where there is a limit on the NH of 5 × 10−4 W/cm3 [10]. For these
Fig. 5. Results for the nuclear heating calculated in the VV inner shell (upper panel) and in the first layer of the internal part of the VV (lower panel).
Fig. 6. Calculated He production in SS- 316 (appm/FPY) at the VV inner shell of the VV for different gap configurations.
two configurations the NH is likely to be above the limit in the TF coil. In order to assess the impact of poloidal gap on the VV heating, the peaking factor has been calculated as the ratio of the NH behind the gap center and the value calculated far from the gap (the angular bin at 10.7◦ ). The peaking factor results to be maximum at the VV inner shell. For 5 cm gap it is 1.4 and it drops to 1.1 only for 2 cm gap. In the first layer of VV behind inner shell the effect is lower. Fig. 6 shows the results of the Helium production in the VV. The limit of the He production for re-weldability is 1 appm [10]. The
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075
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Fig. 7. Calculated damage in DPA/FPY in SS–316 at the inner shell of the VV for different gap configurations.
value of the He production far from the gap ranges between 0.11 and 0.12 appm/FPY. As a consequence, the limit for re-weldability would be reached in 9 FPY. However, major concern arises at position facing the gap center, where the calculated values of the helium production result to be 2.59, 1.54, 1.02 and 0.51 appm/FPY for 5, 3, 2 and 1 cm wide gap, respectively. As a result, the He production is largely influenced by the presence of the gap. This is because it is mostly generated by fast neutrons (E > 0.1 MeV) in steel [11]. It is worth nothing that in the present model there is no Boron content in steel. Boron is very important for He production because it has has a very high impact on it. To estimate the effect of the presence of Boron, assuming 10 ppm (0.001% weight, as recommended for the steel in ITER vacuum vessel cooling tubes), He production behind the blanket has to be multiplied by a factor of 2 [12]. DPA calculation in SS-316 has been performed as well in order to estimate the VV lifetime. In Fig. 7 the DPA at the inner shell of the VV is reported for all gap configurations. The maximum value of the DPA, obtained for the 5 cm gap in front of the gap center is 0.78, with a peaking factor of 1.8. Far from the gap the damage is < 0.5 dpa/FPY. Instead, it results 0.51 and 0.43 dpa/FPY for 2 and 1 cm gap configurations, respectively. Considering the limit of 2.75 dpa on VV steel [10], the 5 cm gap configuration does not meet the requirement, considering 6 FPY of operation. Instead with 1 and 2 cm wide gap the damage limit would be reached in 5.4 and 6.4 years, for 2 and 1 cm wide gaps, respectively. Same calculations were performed using HCPB DEMO homogeneous model for 2 and 5 cm wide gaps. Fig. 8 shows the comparison of the radial neutron flux between the HCLL and HCPB blanket for the case with the 2 cm wide gap, as a function of the radial distance from the first wall (FW). Differences arise even at the FW between the 2 different blankets: the total neutron fluxes result to be 7.3 × 1014 and 4.9 × 1014 n cm−2 s−1 , the fast neutron fluxes result to be 5.5 × 1014 and 3.3 × 1014 n cm−2 s−1 for the HCLL and for the HCPB, respectively. These results are consistent with that reported in [8] for the HCPB, taking into account for the proper normalization. Moreover, the HCPB blanket and manifolds provide a better shielding: the total neutron flux is attenuated by a factor 12.5 behind the blanket and manifold with respect to the flux at the FW with HCPB, compared to an attenuation of 5.4 with HCLL. In particular, for the fast neutron fluxes the attenuation factors are ≈22 with the HCPB blanket and ≈10 for the HCLL one. It should be noted, however, that the blanket model used for HCLL is heterogeneous, whereas the HCPB one is homogeneous and the heterogeneity effect could be not trivial.
Fig. 8. Neutron flux radial profile for HCPB and HCLL DEMO with 2 cm wide poloidal gap. Fast neutron (E > 0.1 MeV) and total neutrons fluxes are shown for both configurations.
Fig. 9. Nuclear heating of the VV as a function of the radial distance from its first wall. Comparison between the HCLL and HCPB blankets with a poloidal gap 2 cm wide.
These results on the neutron fluxes have direct consequences on the nuclear heating of the VV, leading to an overall smaller NH with the HCPB blanket, but with higher peaking factors. In Fig. 9 the mean value of the NH for the 2 cm gap has been shown for both blankets concept. The maximum values is still found for at 10 cm of penetration depth resulting in 0.30 W/cm3 for HCPB, compared with 1.02 W/cm3 for HCLL. The peaking factors are instead larger for HCPB (1.3) with respect to the HCLL. The 5 cm wide gap case have the same trend as the 2 cm wide gap one. The HCPB has a lower NH with respect to the HCLL (both for peak and mean values, 0.36 W/cm3 and 0.33 W/cm3 , respectively) with larger peaking factor (2.2) at the VV first layer. The outcome of this analysis is that the NH of the VV varies significantly for different blanket concepts. Comparing only HCLL and HCPB 2013 blanket configurations, differences on nuclear heating greater than a factor 3 are observed. Also, the peaking factors vary within 35% between the two examined configurations, showing that the peaking factor is strongly dependent on blanket shielding capability. In general it increases as the shielding capability is improved. In ITER, due to better shielding in front of VV, the peaking factor due to the poloidal gap is 3.3 [13]. Moreover, a more complete assessment has also to consider updated blanket concepts. 5. Conclusions The NH nuclear loads of the VV in DEMO HCLL blanket concept has been calculated for different poloidal gap sizes (0, 1, 2, 3 and
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5 cm width) between blanket segments at the inboard equatorial region. The NH has been determined in toroidal and in radial direction. In poloidal direction the effect of the gap has been pointed out, showing a clear peak of the NH corresponding to the center of the gap. The peaking factors on NH in HCLL DEMO are small (within 1.4 with 5 cm gap width): this is because the gamma component in nuclear heating in steel is weakly affected by gap size and because the shielding capability of the 2013HCLL blanket is poor and high nuclear heating is obtained even in zones far from the gap and even without any gap. This study shows that with the HCLL blanket the effect of the poloidal gap could be important even for the TF coil where the limit of 5 × 10−4 W/cm3 could be exceeded with 3 and 5 cm wide gaps. The calculations with HCPB blanket show that the NH of the VV is lower by approximately a factor 3 with respect the HCLL blanket with the same gap configuration. This is due to a better shielding capability of HCPB with respect to HCLL. On the other hand the peaking factors obtained with HCPB are larger. The damage has also been calculated in the VV steel. The maximum value obtained for the 5 cm gap in front of the gap center is 0.78 dpa/FPY, with a peaking factor of 1.8 (5 cm gap). Considering the limit of 2.75 dpa, the only configuration that would not exceed the limit in the DEMO lifetime (∼6 fpy) is the 1 cm wide gap configuration. It may be worthwhile pointing out that for an updated design of the blanket the dpa result to be lower (<0.2 dpa/fpy) [14]. The helium production has been calculated as well. It results to have the largest peaking factors among the calculated quantity, being very sensitive to the gap size. The peak value for the 5 cm wide gap results 2.59 appm per FPY, exceeding the limit of 1 appm for reweldability in 0.39 FPY. Considering a gap size of 2 cm the Heproduction peak is 1.02 appm/FPY and the limit for reweldability is exceeded in ∼1 FPY. However, far from the gap, the he-production is much lower (0.12 appm/FPY) and no problem is foreseen in these zones. It could be noted that the present steel does not content boron which has an high impact on He-production.
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The computing resources and the related technical support used for this work have been provided by CRESCO/ENEAGRID High Performance Computing infrastructure and its staff [15]. CRESCO/ENEAGRID High Performance Computing infrastructure is funded by ENEA, see http://www.cresco.enea.it/english for information. References [1] G. Ramogida, et al., Thermal Structural Analysis Report of Vessel Inner Shell, EUROfusion report, 2014, EFDA D 2D4ZE9. [2] Mcnp, A general Monte Carlo N-particle Trans-port Code, Version 5, Los Alamos National Laboratory, 2013, https://mcnp.lanl.gov/ LA-UR-03-1987. [3] D. Lopez, Aldama, A. Trkov, FENDL-2.1, Update of an Evaluated Nuclear Data Library for Fusion Applications, International Atomic Energy Agency, Vienna, Austria, 2015, 18-July. [4] R. Villari, D. Flammini, F. Moro, Tritium Breeding Ratio Assessment for HCLL DEMO, EFDA D 2ME6F2, 30 May 2014 (https://idm.euro-fusion.org/ ?uid=2ME6F2&action=get document). [5] C. Fausser, et al., Tokamak D-T neutron source models for different plasma physics confinement modes, Fusion Eng. Des. 87 (2012) 787–792. [6] B. Meszaros, Methodology to Derive the Basic DEMO Tokamak ConfigurationCAD Geometry From System Code Studies, Fusion Eng. Design. (2014) (SOFT 2014). [7] C. Bachmann et al., Issues and strategies for DEMO in-vessel component integration, this Volume. [8] P. Pereslavtsev, L. Lu, U. Fischer, O. Bitz, Neutronic analyses of the HCPB DEMO reactor using a consistent integral approach, Fusion Eng. Des. 89 (2014) 1979–1983. [9] T.D. Bohm, M.E. Sawan, B. Smith, P.P.H. Wilson, Peaking in SS Nuclear Parameter at a Water Interface, UWFDM-1376, 2009. [10] U. Fischer, C. Bachmann, I. Palermo, P. Pereslavtsev, R. Villari, Neutronics requirements for a DEMO fusion power plant, in press, Fus. Eng. Des. (2015), http://dx.doi.org/10.1016/j.fusengdes.2015.02.029. [11] M. Shimada, M. Loughlin, M. Shute, Heat and Nuclear Load Specifications for ITER, IDM Number: ITER D 2LULDH v2.3, 16 November 2009, (https://user. iter.org/?uid=2luldh&action=get document). [12] R. Villari, WP12-DTM04-T06: Neutron transport calculation through the DEMO1 blanket (2LNDJ6 v1.1) (February 2013). [13] Heat and Nuclear Load Specifications for ITER, ITER D 2LULDH v2.3 (2009). [14] U. Fischer, et al., Neutronic Performance Issues of Breeder Blanket Options, ISFNT, 2015. [15] G. Ponti, et al., The role of medium size facilities in the HPC ecosystem: the case of the new CRESCO4 cluster integrated in the ENEAGRID infrastructure, Proceedings of the 2014 International Conference on High Performance Computing and Simulation, HPCS (2014) 1030–1033, 6903807.
Acknowledgments This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
Please cite this article in press as: D. Flammini, et al., Neutronics studies for the design of the European DEMO vacuum vessel, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.075