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Detailed 3-D nuclear analysis of ITER blanket modules
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Detailed 3-D nuclear analysis of ITER blanket modules T.D. Bohm a,∗ , M.E. Sawan a , E.P. Marriott a , P.P.H. Wilson a , M. Ulrickson b , J. Bullock b a b
University of Wisconsin-Madison, Madison, WI, USA Sandia National Laboratories, Albuquerque, NM, USA
a r t i c l e
i n f o
Article history: Received 13 September 2013 Received in revised form 16 January 2014 Accepted 16 January 2014 Available online xxx Keywords: MCNP CAD ITER Blanket module
a b s t r a c t In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded. © 2014 Elsevier B.V. All rights reserved.
1. Introduction In ITER, the blanket modules (BMs) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. The BM main components are the first wall (FW) panel and shield block (SB). The FW panel consists of Be armor, Cu heat sink, and steel structure with embedded water coolant. The SB consists of steel structure with embedded water coolant. The numbering convention used to identify the rows of the ITER BMs starts with BM01 on the lower inboard side of the machine, and ends with BM18 on the lower outboard side of the machine. The design process for the BMs includes thermal hydraulics analysis for cooling and thermal stress assessment. Additionally, electromagnetic analysis is required. Re-welding is required in the BM and the VV and this requires accurate determination of helium production. Furthermore, radiation damage in the structural components needs assessment. Therefore, detailed mapping of nuclear heating, He production, and radiation damage is an essential input to the BM design process. Several key ITER components are located adjacent to the BMs such as the edge localized mode (ELM) coils, vertical stabilizer (VS) coils, and blanket manifolds, these also require nuclear analysis. The basic approach we follow for nuclear analysis of a given BM uses a 3-D, “vacuum vessel inward”, 40-degree global CAD model of the ITER machine. This partially simplified model was created
∗ Corresponding author. Tel.: +1 608 262 9312. E-mail address: [email protected] (T.D. Bohm).
for this work from the latest detailed Catia based CAD files of ITER. This model provides the proper nuclear “boundary” environment for detailed BM analysis and is shown in Fig. 1. This model uses partially homogenized components and uses separate models for normal heat flux (NHF) BMs and enhanced heat flux (EHF) BMs. As an example of the level of homogenization used, an NHF BM is composed of 5 separate volumes. BM06 is shown as an example in Fig. 2. Since this simplified 40-degree model is developed for BM analysis, it is called Blanket Lite (BL-Lite). Note that in ITER, the NHF BMs are BM01, BM02, BM06, BM10-BM13, and BM18. The EHF BMs are BM03-BM05, BM07-BM09, and BM14-BM17. When a given BM or other component is ready for nuclear analysis, the homogenized BM is removed and replaced with the detailed CAD model. A volume representing the plasma is used in the BL-Lite model to allow the use of the 3-D ITER sdef based neutron source definition [1]. The final BL-Lite CAD model has 724 volumes and 12,064 surfaces. There are 13 constituent materials in the BL-Lite model and homogenization resulted in 26 different material mixtures. Homogenization of components is performed using the detailed CAD models to determine the appropriate mixed material composition. Comparisons of neutron wall loading (NWL) with other models [2] showed good agreement. For this work, the DAG-MCNP5 Monte Carlo transport code [3], and FENDL-2.1 cross section library [4] were used. Conformal mesh tallies were used to determine detailed maps of nuclear heating in some components. The conformal mesh tallies use a tetrahedral mesh generated with Cubit [5] that conforms to the actual geometry of the given component. This conformal mesh tally feature is particularly useful for components with complex shapes that do not
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Please cite this article in press as: T.D. Bohm, et al., Detailed 3-D nuclear analysis of ITER blanket modules, Fusion Eng. Des. (2014), http://dx.doi.org/10.1016/j.fusengdes.2014.01.056
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Fig. 3. Detailed BM01 model.
Fig. 4. Back view of FW in detailed BM01 model. Fig. 1. Blanket Lite (BL-Lite) CAD model.
A detailed BM01 CAD model was incorporated into the BL-Lite model. The resulting combined model had 993 volumes and 20,360 surfaces. Fig. 3 shows front and back views of the detailed BM01 model. Fig. 4 shows the location of various components at the back of the FW of BM01. Volume averaged nuclear heating was determined in these components. The nuclear heating was 2.0 W/cm3 ,
0.08 W/cm3 , 1.4 W/cm3 , and 0.05 W/cm3 for the upper beam pads, intra-modular keys and pads, lower beam pads, and centering key pads respectively. Statistical uncertainty for these values is less than 5%. Note the lower values of nuclear heating in components near the bottom are due to the lower NWL at that location. Profiles of nuclear heating as a function of depth were also determined for the IM keys, and centering keys in order to allow more detailed thermal engineering analysis of these components. A conformal mesh tally was used to examine the VV nuclear heating behind BM01. Fig. 5 shows the nuclear heating at the front of the VV behind BM01. The statistical uncertainty averages 9% in the individual mesh elements. The nuclear heating values are well below the limit of 0.6 W/cm3 for the VV. Note the structure in the nuclear heating is due to the BM gaps, the cut-outs for the poloidal manifold, and the divertor. The presence of water creates higher nuclear heating in the structure (due to slowing down of neutrons leading to production of more photons, which create most of the nuclear heating in materials like Cu or SS). In addition to the 3-D conformal mesh tallies shown above, a 3D Cartesian mesh tally with voxel sizes of 0.5 cm × 0.5 cm × 0.5 cm was used to determine detailed nuclear heating maps throughout
Fig. 2. Normal heat flux BM (BM06) in the Blanket Lite (BL-Lite) CAD model.
Fig. 5. Nuclear heating in the VV front behind BM01.
match structured meshes. These conformal meshes are also more easily passed to other finite element engineering analysis packages such as computational fluid dynamics or stress analysis. 2. Results A wide variety of nuclear parameters were determined for the analysis of the ITER BMs. For this work, samples of some of the work will be presented here including results for BM01, the neutral beam (NB) region, and the upper port region. 2.1. BM01
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Fig. 6. Nuclear heating in the BM01 region at z = −197 cm.
the whole BM01 region and some of the surrounding region. This particular mesh tally had 7.15 million mesh elements. These types of mesh tally maps are useful in identifying hot spots. Fig. 6 shows a slice through the entire BM01 at z = −197 cm. Fig. 7 shows the location of this slice through the region. Note the structure is clearly visible in Fig. 6 corresponding to water channels in the FW beam, SB, and poloidal manifolds. The statistical uncertainty is <10% in the individual mesh elements in the front half of the BM. Determination of He production in the FW/SB connectors is important as re-welding is required for FW replacement. He production was calculated at the FW/SB connector (bottom of “L” shaped pipe shown in Fig. 4), flow divertor, and branch pipes at ITER’s end of life, 0.54 FPY (1.7 × 107 s). These values were found to be 3.8 appm and 1.7 appm at the front and back of the FW/SB connector, respectively, 2.9 appm and 2.2 appm at the front and back of the flow divertor, respectively, and 0.4 appm at the front of the branch pipes. Statistical uncertainties were ≤9% at these locations. The He production limit for thin welds is 3 appm and since the FW is expected to be replaced before the end of life, these values meet the ITER specification. Note that near SS/water interfaces, the trace amounts of B in the ITER grade SS (10 wppm) creates peaks in He production [6]. However, the B will be depleted over ITER’s lifetime (up to 60%) and therefore these results are conservative as depletion is not taken into account by MCNP [6]. 2.2. NB port Since detailed and clean CAD models were not available, a simplified and partially homogenized model of the NB region was incorporated into the 40-degree BL-Lite model. The combined model had 797 volumes and 13,059 surfaces. Fig. 8 shows a view of the NB region model. Fig. 9 shows a close-up view from the plasma of the NB region model. Note the liners in the heating neutral beam
Fig. 7. Location of slice through the mesh tally in Fig. 6.
Fig. 8. View of the NB region model.
(HNB) port (the larger opening). Fig. 10 shows a view from the VV of the portion of the mid-plane ELM coil that was included in the model. Fig. 11 shows the nuclear heating in CuCrZr for the lower right HNB liner (as viewed from the plasma) determined with a conformal mesh tally. The CuCrZr nuclear heating in the rectangular region near the front averages 7.8 W/cm3 . The statistical uncertainty is 7% in the individual mesh elements shown in the rectangular region. The data in the conformal mesh tally were used to determine a nuclear heating profile along the length of the liner (denoted with an arrow in the figure). The heating profile was fitted to H = 7.8 × exp(−x/91), where H is heating in W/cm3 , and x is the distance along the length of the liner in cm. This heating profile was passed to ITER mechanical designers for further engineering analysis. Nuclear heating and a heating profile were also determined for the left liner. This profile was H = 5.65 × exp(−x/44). The floor of the HNB port is an area of concern since part of SB15 is exposed to the plasma without a first wall panel and because the region is thin, it is difficult to get adequate cooling channels
Fig. 9. View showing detail near the HNB port opening.
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Fig. 13. SS nuclear heating on the VV/port extension with SB15 shown for reference.
Fig. 10. View showing detail of partial ELM coil included in the NB model.
Fig. 14. SS nuclear heating on the lower portion of the VV/port extension near the HNB port.
Fig. 11. CuCrZr nuclear heating in the lower right liner of the HNB port (as viewed from the plasma).
into the region. Fig. 12 shows the nuclear heating in SS for SB15. In the rectangular box near the front of the floor of SB15, the nuclear heating averages 6.7 W/cm3 . The statistical uncertainty averages 5% in the individual mesh elements in this region. Thermal analysts are currently determining if this level of nuclear heating is allowable given the current cooling channel design for SB15. Another area of concern for the NB region is the VV/port extension nuclear heating. The nuclear heating is expected to be highest near the bottom right corner of the HNB port opening (as viewed from the plasma). Fig. 13 shows the SS nuclear heating on the lower part of the VV/port extension with SB15 shown for reference. Fig. 14 shows a close-up view of the SS nuclear heating on the VV/port extension. The nuclear heating averages 1.35 W/cm3 in the rectangular box and the statistical uncertainty averages 6% in the
Fig. 12. SS nuclear heating in SB15.
individual mesh elements. It is clear that a substantial portion of the VV exceeds the 0.6 W/cm3 nuclear heating limit. A possible solution to the excess nuclear heating is utilizing a thicker floor. Based on the peak heating and a typical e-fold distance for the SB material, a design with an 8 cm thicker floor was investigated. Fig. 15 shows a top view of the SS nuclear heating for the VV/port extension with a 0.6 W/cm3 threshold filter applied. The nuclear heating from the original design and the thicker floor design are shown in the figure. It is clear that this thicker floor substantially reduces the amount of the VV/port extension that exceeds the 0.6 W/cm3 limit. There is still a small part of the VV/port extension that exceeds the limit and work is underway to decide on a final design that will meet the VV heating limit. 2.3. Upper port VV heating VV heating near the upper port region is a concern because of the large blanket cut-outs required for the poloidal manifolds, and vertical stabilizer (VS) coils. Because detailed CAD models were not available, a simplified, partially homogenized model of the upper
Fig. 15. SS nuclear heating on the lower portion of the VV/port extension near the HNB port with a 0.6 W/cm3 threshold filter applied. The upper portion shows the original design while the lower portion shows the 8 cm thicker floor design.
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Fig. 16. Close-up view of the upper port model with the VV shells, port extension, VS coils, and poloidal manifolds visible.
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port region. The results show that nuclear heating on the VV and He production on the FW/SB connectors for BM01 are within the specifications. However, the results show that the nuclear heating in the VV/port extension behind the NB region exceeds the limit by a factor of two. The results also show that the nuclear heating in the VV behind the upper port exceeds the limit by 20%. This work is actively being used to refine the design of the components in the regions analyzed. Recall for the analysis of the NB and the upper port regions, partially homogenized models were used. It is expected that the use of detailed models will impact nuclear radiation results at deeper depths. In particular, the use of detailed models will show localized peaks in VV heating which are higher than values reported here. This is primarily because cutouts and gaps between components will be fully resolved (rather than homogenized), allowing radiation to penetrate. The magnitude of localized peaking will depend on the size of the gaps and cutouts. Future work includes analysis of a detailed model for the upper port region (currently undergoing CAD cleaning), analysis of a detailed model of the BM11-13 region to investigate the combined effect due to ELM coil cut-outs, poloidal manifold cut-outs, and the intersection of poloidal and toroidal BM gaps. Additionally, He production in the FW/SB connectors for some OB BMs will be investigated since these have high NWL and the specified limits may be exceeded. Acknowledgments
Fig. 17. SS nuclear heating on the VV inner shell.
port region was used. This upper port model was integrated into a 20-degree version of the BL-Lite model. The combined model had 362 volumes and 7001 surfaces. Fig. 16 shows a close-up view of the upper port region model with only the VV shells, port extension, VS coils, and poloidal manifolds visible. The blankets’ visibility is turned off for clarity. A conformal tetrahedral mesh tally was used to determine the peak nuclear heating of the VV inner shell near the upper port opening. Fig. 17 shows a close-up view of the SS nuclear heating on the VV inner shell. The effect of the toroidal gap and the cut-outs for the VS coil and poloidal manifolds is readily apparent. The average heating in the large rectangular region is 0.73 W/cm3 and the average heating in the small rectangular region is 0.67 W/cm3 . The statistical uncertainty is 6% in the individual mesh elements in these regions. Note there is a substantial portion of the VV inner shell with nuclear heating greater than the 0.6 W/cm3 limit. 3. Conclusions and future work A 40-degree global CAD model of ITER was used as the basis for 3-D based nuclear analysis of BM01, the NB region, and the upper
The support for this work has been provided by Sandia National Laboratory. Sandia is a multi-program laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. Part of this work was carried out using a portion of the B-lite MCNP model which was developed as a collaborative effort between the FDS team of ASIPP China, University of Wisconsin-Madison, ENEA Frascati, CCFE UK, JAEA Naka, and the ITER Organization. References [1] E. Polunovskiy, SDEF card for the ITER standard neutron source inductive operation scenario with 500 MW of fusion power, IDM ITER D 2KS8CN v1.4, 2010. [2] P.P.H. Wilson, R. Feder, U. Fischer, M. Loughlin, L. Petrizzi, Y. Wu, et al., State of the art 3-D radiation transport methods for fusion energy systems, Fusion Engineering and Design 83 (2008) 824–833. [3] T. Tautges, P.P.H. Wilson, J. Kraftcheck, B. Smith, D. Henderson, Acceleration techniques for direct use of cad-based geometries in Monte Carlo radiation transport, in: Proceedings of International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2009), Saratoga Springs, New York, May 3–7, 2009. [4] D.L. Aldama, A. Trkov, FENDL-2.1, Update of an Evaluated Nuclear Data Library for Fusion Applications, Report INDC(NDS)-467, International Atomic Energy Agency, 2004. [5] B. Hanks, CUBIT Geometry and Mesh Generation Toolkit, version 13.2, Computation Simulation Infrastructure, Sandia National Laboratories, 2012. [6] T.D. Bohm, M.E. Sawan, B. Smith, P.P.H. Wilson, Investigation of observed peaking in nuclear parameters at steel/water interfaces, Fusion Science and Technology 60 (2011) 698–702.
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FUSION-7233; No. of Pages 5
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Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Detailed 3-D nuclear analysis of ITER blanket modules T.D. Bohm a,∗ , M.E. Sawan a , E.P. Marriott a , P.P.H. Wilson a , M. Ulrickson b , J. Bullock b a b
University of Wisconsin-Madison, Madison, WI, USA Sandia National Laboratories, Albuquerque, NM, USA
a r t i c l e
i n f o
Article history: Received 13 September 2013 Received in revised form 16 January 2014 Accepted 16 January 2014 Available online xxx Keywords: MCNP CAD ITER Blanket module
a b s t r a c t In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded. © 2014 Elsevier B.V. All rights reserved.
1. Introduction In ITER, the blanket modules (BMs) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. The BM main components are the first wall (FW) panel and shield block (SB). The FW panel consists of Be armor, Cu heat sink, and steel structure with embedded water coolant. The SB consists of steel structure with embedded water coolant. The numbering convention used to identify the rows of the ITER BMs starts with BM01 on the lower inboard side of the machine, and ends with BM18 on the lower outboard side of the machine. The design process for the BMs includes thermal hydraulics analysis for cooling and thermal stress assessment. Additionally, electromagnetic analysis is required. Re-welding is required in the BM and the VV and this requires accurate determination of helium production. Furthermore, radiation damage in the structural components needs assessment. Therefore, detailed mapping of nuclear heating, He production, and radiation damage is an essential input to the BM design process. Several key ITER components are located adjacent to the BMs such as the edge localized mode (ELM) coils, vertical stabilizer (VS) coils, and blanket manifolds, these also require nuclear analysis. The basic approach we follow for nuclear analysis of a given BM uses a 3-D, “vacuum vessel inward”, 40-degree global CAD model of the ITER machine. This partially simplified model was created
∗ Corresponding author. Tel.: +1 608 262 9312. E-mail address: [email protected] (T.D. Bohm).
for this work from the latest detailed Catia based CAD files of ITER. This model provides the proper nuclear “boundary” environment for detailed BM analysis and is shown in Fig. 1. This model uses partially homogenized components and uses separate models for normal heat flux (NHF) BMs and enhanced heat flux (EHF) BMs. As an example of the level of homogenization used, an NHF BM is composed of 5 separate volumes. BM06 is shown as an example in Fig. 2. Since this simplified 40-degree model is developed for BM analysis, it is called Blanket Lite (BL-Lite). Note that in ITER, the NHF BMs are BM01, BM02, BM06, BM10-BM13, and BM18. The EHF BMs are BM03-BM05, BM07-BM09, and BM14-BM17. When a given BM or other component is ready for nuclear analysis, the homogenized BM is removed and replaced with the detailed CAD model. A volume representing the plasma is used in the BL-Lite model to allow the use of the 3-D ITER sdef based neutron source definition [1]. The final BL-Lite CAD model has 724 volumes and 12,064 surfaces. There are 13 constituent materials in the BL-Lite model and homogenization resulted in 26 different material mixtures. Homogenization of components is performed using the detailed CAD models to determine the appropriate mixed material composition. Comparisons of neutron wall loading (NWL) with other models [2] showed good agreement. For this work, the DAG-MCNP5 Monte Carlo transport code [3], and FENDL-2.1 cross section library [4] were used. Conformal mesh tallies were used to determine detailed maps of nuclear heating in some components. The conformal mesh tallies use a tetrahedral mesh generated with Cubit [5] that conforms to the actual geometry of the given component. This conformal mesh tally feature is particularly useful for components with complex shapes that do not
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Please cite this article in press as: T.D. Bohm, et al., Detailed 3-D nuclear analysis of ITER blanket modules, Fusion Eng. Des. (2014), http://dx.doi.org/10.1016/j.fusengdes.2014.01.056
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Fig. 3. Detailed BM01 model.
Fig. 4. Back view of FW in detailed BM01 model. Fig. 1. Blanket Lite (BL-Lite) CAD model.
A detailed BM01 CAD model was incorporated into the BL-Lite model. The resulting combined model had 993 volumes and 20,360 surfaces. Fig. 3 shows front and back views of the detailed BM01 model. Fig. 4 shows the location of various components at the back of the FW of BM01. Volume averaged nuclear heating was determined in these components. The nuclear heating was 2.0 W/cm3 ,
0.08 W/cm3 , 1.4 W/cm3 , and 0.05 W/cm3 for the upper beam pads, intra-modular keys and pads, lower beam pads, and centering key pads respectively. Statistical uncertainty for these values is less than 5%. Note the lower values of nuclear heating in components near the bottom are due to the lower NWL at that location. Profiles of nuclear heating as a function of depth were also determined for the IM keys, and centering keys in order to allow more detailed thermal engineering analysis of these components. A conformal mesh tally was used to examine the VV nuclear heating behind BM01. Fig. 5 shows the nuclear heating at the front of the VV behind BM01. The statistical uncertainty averages 9% in the individual mesh elements. The nuclear heating values are well below the limit of 0.6 W/cm3 for the VV. Note the structure in the nuclear heating is due to the BM gaps, the cut-outs for the poloidal manifold, and the divertor. The presence of water creates higher nuclear heating in the structure (due to slowing down of neutrons leading to production of more photons, which create most of the nuclear heating in materials like Cu or SS). In addition to the 3-D conformal mesh tallies shown above, a 3D Cartesian mesh tally with voxel sizes of 0.5 cm × 0.5 cm × 0.5 cm was used to determine detailed nuclear heating maps throughout
Fig. 2. Normal heat flux BM (BM06) in the Blanket Lite (BL-Lite) CAD model.
Fig. 5. Nuclear heating in the VV front behind BM01.
match structured meshes. These conformal meshes are also more easily passed to other finite element engineering analysis packages such as computational fluid dynamics or stress analysis. 2. Results A wide variety of nuclear parameters were determined for the analysis of the ITER BMs. For this work, samples of some of the work will be presented here including results for BM01, the neutral beam (NB) region, and the upper port region. 2.1. BM01
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Fig. 6. Nuclear heating in the BM01 region at z = −197 cm.
the whole BM01 region and some of the surrounding region. This particular mesh tally had 7.15 million mesh elements. These types of mesh tally maps are useful in identifying hot spots. Fig. 6 shows a slice through the entire BM01 at z = −197 cm. Fig. 7 shows the location of this slice through the region. Note the structure is clearly visible in Fig. 6 corresponding to water channels in the FW beam, SB, and poloidal manifolds. The statistical uncertainty is <10% in the individual mesh elements in the front half of the BM. Determination of He production in the FW/SB connectors is important as re-welding is required for FW replacement. He production was calculated at the FW/SB connector (bottom of “L” shaped pipe shown in Fig. 4), flow divertor, and branch pipes at ITER’s end of life, 0.54 FPY (1.7 × 107 s). These values were found to be 3.8 appm and 1.7 appm at the front and back of the FW/SB connector, respectively, 2.9 appm and 2.2 appm at the front and back of the flow divertor, respectively, and 0.4 appm at the front of the branch pipes. Statistical uncertainties were ≤9% at these locations. The He production limit for thin welds is 3 appm and since the FW is expected to be replaced before the end of life, these values meet the ITER specification. Note that near SS/water interfaces, the trace amounts of B in the ITER grade SS (10 wppm) creates peaks in He production [6]. However, the B will be depleted over ITER’s lifetime (up to 60%) and therefore these results are conservative as depletion is not taken into account by MCNP [6]. 2.2. NB port Since detailed and clean CAD models were not available, a simplified and partially homogenized model of the NB region was incorporated into the 40-degree BL-Lite model. The combined model had 797 volumes and 13,059 surfaces. Fig. 8 shows a view of the NB region model. Fig. 9 shows a close-up view from the plasma of the NB region model. Note the liners in the heating neutral beam
Fig. 7. Location of slice through the mesh tally in Fig. 6.
Fig. 8. View of the NB region model.
(HNB) port (the larger opening). Fig. 10 shows a view from the VV of the portion of the mid-plane ELM coil that was included in the model. Fig. 11 shows the nuclear heating in CuCrZr for the lower right HNB liner (as viewed from the plasma) determined with a conformal mesh tally. The CuCrZr nuclear heating in the rectangular region near the front averages 7.8 W/cm3 . The statistical uncertainty is 7% in the individual mesh elements shown in the rectangular region. The data in the conformal mesh tally were used to determine a nuclear heating profile along the length of the liner (denoted with an arrow in the figure). The heating profile was fitted to H = 7.8 × exp(−x/91), where H is heating in W/cm3 , and x is the distance along the length of the liner in cm. This heating profile was passed to ITER mechanical designers for further engineering analysis. Nuclear heating and a heating profile were also determined for the left liner. This profile was H = 5.65 × exp(−x/44). The floor of the HNB port is an area of concern since part of SB15 is exposed to the plasma without a first wall panel and because the region is thin, it is difficult to get adequate cooling channels
Fig. 9. View showing detail near the HNB port opening.
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Fig. 13. SS nuclear heating on the VV/port extension with SB15 shown for reference.
Fig. 10. View showing detail of partial ELM coil included in the NB model.
Fig. 14. SS nuclear heating on the lower portion of the VV/port extension near the HNB port.
Fig. 11. CuCrZr nuclear heating in the lower right liner of the HNB port (as viewed from the plasma).
into the region. Fig. 12 shows the nuclear heating in SS for SB15. In the rectangular box near the front of the floor of SB15, the nuclear heating averages 6.7 W/cm3 . The statistical uncertainty averages 5% in the individual mesh elements in this region. Thermal analysts are currently determining if this level of nuclear heating is allowable given the current cooling channel design for SB15. Another area of concern for the NB region is the VV/port extension nuclear heating. The nuclear heating is expected to be highest near the bottom right corner of the HNB port opening (as viewed from the plasma). Fig. 13 shows the SS nuclear heating on the lower part of the VV/port extension with SB15 shown for reference. Fig. 14 shows a close-up view of the SS nuclear heating on the VV/port extension. The nuclear heating averages 1.35 W/cm3 in the rectangular box and the statistical uncertainty averages 6% in the
Fig. 12. SS nuclear heating in SB15.
individual mesh elements. It is clear that a substantial portion of the VV exceeds the 0.6 W/cm3 nuclear heating limit. A possible solution to the excess nuclear heating is utilizing a thicker floor. Based on the peak heating and a typical e-fold distance for the SB material, a design with an 8 cm thicker floor was investigated. Fig. 15 shows a top view of the SS nuclear heating for the VV/port extension with a 0.6 W/cm3 threshold filter applied. The nuclear heating from the original design and the thicker floor design are shown in the figure. It is clear that this thicker floor substantially reduces the amount of the VV/port extension that exceeds the 0.6 W/cm3 limit. There is still a small part of the VV/port extension that exceeds the limit and work is underway to decide on a final design that will meet the VV heating limit. 2.3. Upper port VV heating VV heating near the upper port region is a concern because of the large blanket cut-outs required for the poloidal manifolds, and vertical stabilizer (VS) coils. Because detailed CAD models were not available, a simplified, partially homogenized model of the upper
Fig. 15. SS nuclear heating on the lower portion of the VV/port extension near the HNB port with a 0.6 W/cm3 threshold filter applied. The upper portion shows the original design while the lower portion shows the 8 cm thicker floor design.
Please cite this article in press as: T.D. Bohm, et al., Detailed 3-D nuclear analysis of ITER blanket modules, Fusion Eng. Des. (2014), http://dx.doi.org/10.1016/j.fusengdes.2014.01.056
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ARTICLE IN PRESS T.D. Bohm et al. / Fusion Engineering and Design xxx (2014) xxx–xxx
Fig. 16. Close-up view of the upper port model with the VV shells, port extension, VS coils, and poloidal manifolds visible.
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port region. The results show that nuclear heating on the VV and He production on the FW/SB connectors for BM01 are within the specifications. However, the results show that the nuclear heating in the VV/port extension behind the NB region exceeds the limit by a factor of two. The results also show that the nuclear heating in the VV behind the upper port exceeds the limit by 20%. This work is actively being used to refine the design of the components in the regions analyzed. Recall for the analysis of the NB and the upper port regions, partially homogenized models were used. It is expected that the use of detailed models will impact nuclear radiation results at deeper depths. In particular, the use of detailed models will show localized peaks in VV heating which are higher than values reported here. This is primarily because cutouts and gaps between components will be fully resolved (rather than homogenized), allowing radiation to penetrate. The magnitude of localized peaking will depend on the size of the gaps and cutouts. Future work includes analysis of a detailed model for the upper port region (currently undergoing CAD cleaning), analysis of a detailed model of the BM11-13 region to investigate the combined effect due to ELM coil cut-outs, poloidal manifold cut-outs, and the intersection of poloidal and toroidal BM gaps. Additionally, He production in the FW/SB connectors for some OB BMs will be investigated since these have high NWL and the specified limits may be exceeded. Acknowledgments
Fig. 17. SS nuclear heating on the VV inner shell.
port region was used. This upper port model was integrated into a 20-degree version of the BL-Lite model. The combined model had 362 volumes and 7001 surfaces. Fig. 16 shows a close-up view of the upper port region model with only the VV shells, port extension, VS coils, and poloidal manifolds visible. The blankets’ visibility is turned off for clarity. A conformal tetrahedral mesh tally was used to determine the peak nuclear heating of the VV inner shell near the upper port opening. Fig. 17 shows a close-up view of the SS nuclear heating on the VV inner shell. The effect of the toroidal gap and the cut-outs for the VS coil and poloidal manifolds is readily apparent. The average heating in the large rectangular region is 0.73 W/cm3 and the average heating in the small rectangular region is 0.67 W/cm3 . The statistical uncertainty is 6% in the individual mesh elements in these regions. Note there is a substantial portion of the VV inner shell with nuclear heating greater than the 0.6 W/cm3 limit. 3. Conclusions and future work A 40-degree global CAD model of ITER was used as the basis for 3-D based nuclear analysis of BM01, the NB region, and the upper
The support for this work has been provided by Sandia National Laboratory. Sandia is a multi-program laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. Part of this work was carried out using a portion of the B-lite MCNP model which was developed as a collaborative effort between the FDS team of ASIPP China, University of Wisconsin-Madison, ENEA Frascati, CCFE UK, JAEA Naka, and the ITER Organization. References [1] E. Polunovskiy, SDEF card for the ITER standard neutron source inductive operation scenario with 500 MW of fusion power, IDM ITER D 2KS8CN v1.4, 2010. [2] P.P.H. Wilson, R. Feder, U. Fischer, M. Loughlin, L. Petrizzi, Y. Wu, et al., State of the art 3-D radiation transport methods for fusion energy systems, Fusion Engineering and Design 83 (2008) 824–833. [3] T. Tautges, P.P.H. Wilson, J. Kraftcheck, B. Smith, D. Henderson, Acceleration techniques for direct use of cad-based geometries in Monte Carlo radiation transport, in: Proceedings of International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2009), Saratoga Springs, New York, May 3–7, 2009. [4] D.L. Aldama, A. Trkov, FENDL-2.1, Update of an Evaluated Nuclear Data Library for Fusion Applications, Report INDC(NDS)-467, International Atomic Energy Agency, 2004. [5] B. Hanks, CUBIT Geometry and Mesh Generation Toolkit, version 13.2, Computation Simulation Infrastructure, Sandia National Laboratories, 2012. [6] T.D. Bohm, M.E. Sawan, B. Smith, P.P.H. Wilson, Investigation of observed peaking in nuclear parameters at steel/water interfaces, Fusion Science and Technology 60 (2011) 698–702.
Please cite this article in press as: T.D. Bohm, et al., Detailed 3-D nuclear analysis of ITER blanket modules, Fusion Eng. Des. (2014), http://dx.doi.org/10.1016/j.fusengdes.2014.01.056