Neutronics analysis of inboard shielding capability for a DEMO fusion reactor

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1 *Manuscript Click here to view linked References Neutronics analysis of inboard shielding capability for a DEMO fusion reactor Songlin Liu a, Jiangang Li a, Shanliang Zheng b, Neil Mitchell c a Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, , China b Culham Centre for Fusion Energy, Abingdon, Oxfordshire, OX14 3DB, UK c ITER Organization, Route de Vinon sur Verdon, Saint Paul Lez Durance, France Abstract: The inboard shielding for the fusion reactor can be a crucial issue due to the limited space available in a tokamak configuration. It is necessary to assess the inboard shielding capability of DEMO for its initial design. In this paper, one-dimensional (1D) and three-dimensional (3D) neutronics models were developed based on a reference design of the Chinese Fusion Engineering Testing Reactor (CFETR). In this DEMO reactor, the neutron wall load (NWL) is in the range of 1.5~3 MW/m 2 and the inboard shielding thickness are constrained within 40cm - 70cm which reserves enough room for tritium breeding to achieve the tritium self-sufficiency of the reactor. Referring to the detailed design of the ITER Toroidal Field Coils (TFC) and using radiation hardening technology developed for ITER, the inboard blanket shielding capability and nuclear responses of the TFC are investigated for both FLiBe and Li 4 SiO 4 breeding blanket concepts. Some scoping calculations for the inboard shielding are carried out using 1D model. Tritium Breeding Ratio (TBR) is validated in 3D model to assure the tritium self-sufficiency and the impact of the gaps on shielding performance is discussed. Some suggestions on improving the inboard shielding performance for DEMO are also proposed. Keywords: fusion, DEMO, radiation shielding, tritium breeding 1. Introduction One of key roles of a fusion DEMO reactor is to demonstrate energy production and its tritium selfsufficiency with installing a suitable breeding blanket. For a superconducting tokamak fusion DEMO reactor, the in-vessel shielding system, comprised of the tritium breeding blanket (BB) and the shielding blanket (SB), and the vacuum vessel (VV), needs to provide adequate shielding which protects the TFCs from being damaged by 14.06MeV D-T fusion neutrons. Particularly, in inboard direction of such a reactor, shielding is most crucial and is a concern for the fusion community due to the limited space available [1-2]. The shielding capability and thickness of the shield required for a fusion reactor depends on operating plasma parameters and both physics and engineering needs, including the neutron wall load (NWL) and neutron fluence, shielding materials composition and the radiation load limit to TFCs [3]. This paper presents the investigation of the impact of various tritium breeders and shielding materials on the inboard shielding capability and the corresponding thickness with the guarantee of the tritium self-sufficiency. Taking the typical DEMO NWL range, 1.5~3 MW/m 2, the calculations are performed using an one-dimensional (1D) model comparing with the radiation limits of the ITER TFCs. The original radiation limits of the coils as given in [4], i.e. limits of fast neutron (>0.1MeV) fluence in Winding Pack (WP) magnet superconductor n/cm 2, and in WP insulator n/cm 2, limits of peak local nuclear heating in magnet steel case W/cm 3 and in magnet conductor W/cm 3 have been modified in the course of the design. In particular the use of cyanate ester resins as the base filler for the coil insulation system allows neutron fluences up to n/cm 2, a factor of 4 better than the original design criteria [5-6] based on epoxy resin filler. On the other hand, the ITER coils can tolerate peak local nuclear heating in magnet steel case of W/cm 3 and in magnet conductor W/cm 3 because of the integral heat removal capability of the design [7], about 1/3 of the original criteria. The ITER TF coil heating limit is 14kW [7] but to scale to CFETR a number of factors must be considered. Of the 14kW, 10.9kW is deposited in the inner legs of the TFC, the subject of the present study. The heat removal is limited by the allowable temperature rise of the Helium in the inner leg region, about 0.3K, which is in turn linked to the peak field on the conductor. The peak field in ITER is 11.8T, compared to about 10.3T in CFETR. This field reduction greatly increases the temperature margin of the superconductor (by over 1K). If required, there is therefore scope to increase the ITER case and coil local heating limits for CFETR, by over a factor of 2, with essentially the same design. This will however result in a higher cryoplant load and although it is technically feasible, it is commercially undesirable.

2 A three-dimensional (3D) calculation is also carried out to validate tritium production capability of this design and the impact of the gaps on the shielding requirement. In this paper, Section 2 describes the calculation models and tools; the calculation results and analyses are presented in Section 3; the discussion and summary are given in Section 4 and Section 5, respectively. 2. Models and tools Taking one set of primary parameters of a reference design Chinese Fusion Engineering Testing Reactor (CFETR) [8], the major and minor radii are R=5.5m and a= 1.5m respectively, as a model base, 1D cylindrical neutronics model, with the height of 1 m and the top and bottom planes are both reflectory boundaries, and 3D model of 20 degree are developed (shown in Fig.1), which contain all the representative inboard and outboard components, such as the plasma chamber, first wall (FW), breeder blanket (BB), shield blanket (SB), vacuum vessel (VV), TFC case, TFC conductor, and TFC insulator. The neutron source normalised by the NWL proposed in the reference design so that the estimation of 1D model accommodates the maximum neutron flux and focuses on the scoping calculation for any further optimization. For 3D model, the 2cm gaps between blanket modules along both toroidal and poloidal directions are adopted. The initial radial thickness and materials composition of VV and TFC at inboard side of both sets of models are same as ITER s, and is the materials for each component are homogenised to simplify the model. (a) Fig.1 Neutronis calculation model: (a) 1D cylindrical model, (b) 3D model of 20 degree. Table 1: Material composition (Vol.%) of breeding blanket FLiBe Blanket FW RAFM, Steel:75% Helium gas: 24% Breeding zone RAFM steel, 20% Breeder:FLiBe 25%, Li6 enrichment 60% Multiplier:Be 55% (b) Li 4SiO 4 blanket RAFM steel, 75% Helium gas: 24% RAFM steel, 20% Breeder:Li 4SiO 4 15% Packing factor 0.62, Li6 enrichment 80% Multiplier:Be65%, Packing factor 0.8 Two breeding blanket concepts, self-cooled molten salt FLiBe blanket and Helium-cooled Li 4 SiO 4 ceramic pebble bed blanket, are selected to investigate the impact on shielding requirements by different tritium breeders. The materials composition of two type breeding blankets are summarised in table 1. Prior to the validation on 3D model, the shielding thickness, materials compositions and the enrichment of Li-6 in the tritium breeding blanket have been optimized based on a 1D sphere model which are not given in details in this paper. As one part of shielding components, SB has to function as an efficient neutron moderator, such as a hydrogenous material, and a good neutron absorber, such as steel, boronated steel, tungsten, tungsten carbide and boron carbide. Different shielding materials present different shielding performances and costs. In this study, three shield blanket variants are taken into account, 1) case 1: 60%SS; 40%H2O, 2) case2: 84.4%SS; 14.3%H2O, and 3) case3: 75%WC; 15%H2O; 10%SS (these are volume percentages). Like ITER shielding blanket, case1 and case2 employ mature-technology and relatively cheap materials, and the advanced material, WC, is applied in case 3. For three cases, the thickness of shielding blanket is optimized at the range of 40~70 cm. A series of nuclear responses of different models are calculated using the MCNP/4C code [9] and the IAEA Fusion Evaluated Nuclear Data Library FENDL2.1 [10]. 3. Calculation and analysis The TBR of 1.58 for Li 4 SiO 4 blanket and 1.62 for FLiBe blanket are achieved in the 1D models where the inboard and outboard breeding blanket zones are 30cm and 70cm, respectively. The models are subsequently employed to the inboard shielding efficiency study. As suggested in ref. [11], considering the uncertainties of 1D analyses, the 3D correction factors in ITER TFC inboard leg neutronic analysis are employed in present 1D study as conservative estimation, i.e. gap effect factor of 2, fexible joint factor of 1.2, and peaking factor in inboard of 1.37 are applied to correct the results from 1D calculation. Although the DEMO blanket, different from ITER shielding blanket, contains both breeding and shielding blankets and may employ different candidate shielding materials, at the current stage of the nuclear analysis, the 3D correction factors recommended by ITER are still effectively valuable before a detailed 3D model is developed for any DEMO reactor. 3.1 Shielding requirements for different breeders For both Li 4 SiO 4 and FLiBe breeder blankets, Fig.2 shows the neutron spectrum on the back plate of breeding blanket corresponding to the NWL of 1.5MW/m 2. It is observed that the neutrons escaping away from FLiBe breeder blanket are less than those from Li 4 SiO 4 breeder blanket. After 14.1MeV neutrons produced in D-T reaction in fusion plasma migrate through FW and BB into SB, most neutrons have been moderated by various materials in those components. This indicates that increasing the fraction of heavy nuclides in shielding materials may provide better shielding due to the

3 inelastic scattering and absorption neutron reactions. Fig.3 illustrates the fast neutron (>0.1MeV) flux density radial profile for case 1 with the inboard SB ~70cm in thickness and it is normalised to the NWL ~1.5MW/m 2. It is demonstrated that the shielding efficiency of the FLiBe breeder blanket is relatively higher for fast neutrons. Fig.2 The neutron spectrum on back plate of both FliBe and Li 4 SiO 4 breeding blanket under 1.5 MW/m 2 NWL. Fig.3 The fast neutron flux radial profile at inboard mid plane under 1.5MW/m 2 NWL in case Radiation Loads to TFC The irradiation scenarios for the CFETR are preliminarily designed as 10 and 20 FYP (full power year) with the assumption that it will operate 20 years and 40 years with a duty fact of 50%. A series of nuclear responses of TFC are calculated for the Li 4 SiO 4 breeder blanket corresponding to the NWL of 1.5~3MW/m 2 varying the thickness of SB with different shielding materials of cases 1, 2 and 3 respectively. Fig.4 illustrates fast neutron (>0.1MeV) fluence in WP s insulator as a function of the thickness of shielding blanket and the original and new fast neutron fluence limits to the TFC insulation. All other nuclear responses, such as the fast neutron fluence in the WP, peak local nuclear heat density in magnet steel case, and peak local nuclear heat density in magnet conductor, meet the original limits for ITER TFC given in section 1. Based on the results of Fig.4, the minimum thicknesses of shielding blankets to provide adequate shielding are extrapolated and summarized in Table 2. It s shown that case 3, the mixture of 75%WC, 15%H 2 O and 10%SS as the shielding materials in SB, is able to provide the strongest shielding capability for TFC assuming the same SB thickness. 3.3 Effects of 3D model Generally, it is effective and less time consuming to use 1D neutronics model to perform preliminary calculations and optimizations. However, in a fusion reactor, the radiation streaming through the numerous gaps, slots, and any other major penetrations in in-vessel components will certainly have impact on the shielding capability. In this study, 3D neutronics model is developed to carry out the analysis taking the gaps into account and to validate the results obtained from the 1D model. The TBR of 1.17 in 3D model for Li4SiO4 breeder blanket indicates this breeding blanket meets the requirement of the tritium self-sufficiency with some margin to allow some loss during the fuel cycling. The TBR decreases ~26% in 3D model comparing with 1D result because the gaps and divertor region are taken into account in 3D model which reduce the effective tritium breeding volumes. (a) (b) (c) Fig.4 Fast neutron (>0.1MeV) fluence in WP s insulator vs. radial thickness of SB in (a) case1; (b)case2; (c) case3. Dotted lines show ITER improved limit and original design criteria

4 Table 2 The minimum thickness of shielding blanket and the heating density to TFC NWL ITEM Case1 Case2 Case3 10 FPY 20FPY 10 FPY 20FPY 10 FPY 20FPY 1.5MW/m 2 Thickness, cm Heating a), MW/m E E E E E E-5 Heating b), MW/m E E E E E E-5 2.5MW/m 2 Thickness, cm c) c) Heating a), MW/m E E E E-5 Heating b), MW/m E E E E-5 3.0MW/m 2 Thickness, cm c) c) Heating a), MW/m E E E E-5 Heating b), MW/m E E E E-6 a) Peak nuclear heating density in magnet steel case; b)peak nuclear heating density in magnet conductor; c)extrapolation value. For the SB of case 1 with 70 cm radial thickness, the nuclear responses to TFC at mid-plane are calculated using 3D model and listed in Table 3 together with 1D result normalized to the NWL of 1.5MW/m 2. Comparing with the 1D results re-evaluated using the ITER 3D correction factor, fast neutron fluence increases ~40% in WP magnet superconductor and ~10% in WP insulator; Table 3 Comparison of the nuclear load to TFC in 1D analysis with 3D s however, nuclear heating in magnet steel case and in magnet conductor decreases ~35%. All the nuclear responses of TFC are satisfactory in terms of the ITER design limits. This means the application of ITER 3D correction factor in 1D calculation is applicable as the approximate estimation for the DEMO reactor. Response Original Criteria ITER Design Limit Case1,70cm thickness,10fpy 1D result 1D result a) 3D result Fast neutron fluence in Winding Pack (WP) 1 10 magnet superconductor n/cm n/cm E E E+17 Fast neutron fluence in WP insulator n/cm n/cm E E E+17 Nuclear heating in magnet steel case W/cm W/cm E E E-05 Nuclear heating in magnet conductor W/cm W/cm E E E-05 a) Using the ITER 3D correction factor 4. Discussion The shielding capability of the SB with the same thickness and NWL varies with the types of the shielding materials and the ratio to water. The shielding materials of case 1 and case 2, similar to those used in ITER, may meet the shielding needs at the cost of the thickness; however, it might come into conflict between the allocations of the central solenoid coils and of the inboard blanket which means a larger reactor configuration may be required in order to accommodate all the needs, including tritium production, shielding adequacy and essential components alloactions etc. As an advanced material candidate, case 3 demonstrates the strongest shielding performance with the minimum thickness although the cost of WC is expensive and it is hard to achieve its theoretical density in industry fabrication. But if the limited inboard space sits on the top issue, this advanced material might be taken as main candidate for inboard shield design of DEMO. In this work, as an advanced shielding material, only WC is investigated so far. According to reference [12], B 4 C or H 2 Zr may act as good candidate shielding materials for future DEMO. It is observed that the breeding blanket may moderate most neutrons to low and intermediary energy due to various neutron induced reactions. The current design of the blanket scheme is still very rough. It may leave some room to further reduce neutron leakage and enhance shielding capability by considering carefully the combination of the breeding blanket design and the neutron reflector in blanket. In terms of nuclear responses in TFC, although they all appear satisfactory to the ITER TFC design limits, the fast neutron fluence in WP insulator is more serious than the integrated heat which may be accommodated by the cooling system design. For future DEMO reactors, already use of the advanced insulation technology developed for ITER [4-5] brings a substantial advantage, reducing the overall blanket thickness by about 10cms with fluence up to n/сm 2 while offering the same mechanical performance as conventional systems. Further improvements can perhaps be made, such as using advanced insulator material with further technology development, for example, nonsilicate ceramic (Al 2 O 3, MgO) with fluence to (0.5 1) n/сm 2 and TIN-1 (expected fluence n/сm 2 ). [ 13 ] However such material is mechanically weak and requires a new TFC design. 5. Summary In this work, 1D and 3D neutronics models have been developed based on a reference design of the CFETR. Both FLiBe and Li 4 SiO 4 breeding concepts and three sets of shielding materials compositions were selected to

5 perform the nuclear analysis taking the design limits of ITER TFC. The inboard blanket shielding capability has been assessed under the condition of the NWL ~1.5-3 MW/m 2 with the thickness constraints of 40~70 cm using 1D model to perform the scoping calculations. Preliminary results indicate the total inboard blanket thickness of ~100 cm (BB ~30 cm, SB ~60-80 cm) can provide the adequate shielding at the NWL range of 1.5~3 WM/m 2 while using the mixture of SS and water as the shielding material. If the SB adopts some advanced shielding material, such as tungsten carbide mixed with water as the coolant and steel as the structure wall, the thickness of SB can be reduced to occupy less space. Different breeders make little impact on shielding towards TFC; but there is still some room and possibility for further optimization. The use of the radiation hardened insulator developed for the ITER TFC brings a gain of almost 10cm in blanket thickness reduction, itself leading to further improvements in CFETR due to a reduction in TFC magnetic field and more space for the Central Solenoid. TFC heating is clearly less critical than the fluence limits and margins are available. In the 3D model developed in this work, only gaps between the blanket modules have been taken into account but any other possible streaming channels in the components have been neglected. Therefore, the model with heterogeneous components needs to be further developed in order to perform more accurate analyses. Acknowledgements This work was supported by Chinese the National Natural Science Foundation under Grant No , Reference [1] J. Jordanova, et al. Fus. Eng.&Des., 81(19): [2] T. Hayashi, et al. Fus. Eng.&Des., 81(8-14): [3] R. T. Santoro. Radiation Shielding for Fusion Reactors. Joural of nuclear science and technology. [4] Nuclear Shielding for TF coils, ITER_D_4ABRCH. [5] K. Humer, et al., Fus. Eng.&Des., 84(2-6)2009. [6] F. Savary, et al., presented at MT22 and published in IEEE Trans Applied Superconductivity, [ 7 ] ITER Magnet Design Description Document Part 1, ITER_D_2NPLKM, version 1.8, Sept [ 8 ] Jiangang Li, Hefei, China, May [9] J.F. Briesmeister (Ed.), Report LA M, April [10] IAEA FENDL-2.1, INDC (NDS)-467, December [11] ITER NAR, July 2004, ITER_D_22F2ST. [12] Y. CHEN, et al., FZKA 6763 (April 2003). [ 13 ] E. Azizov, Hefei, China, May 2012.