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Nuclear Technology

Nuclear Technology. Andrew Rosenstrom, Joe Elkins, Rahat Ahmed, Justin Alexander. Reference Nuclear Technology Parameters for International DEMO Designs. Source: [2],[3],[10],[11],[12],[13],[14],[15]. Guiding Nuclear Technology Parameters for Georgia Tech DEMO Design. Stress Calculations.

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Nuclear Technology

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  1. Nuclear Technology Andrew Rosenstrom, Joe Elkins, Rahat Ahmed, Justin Alexander

  2. Reference Nuclear Technology Parameters for International DEMO Designs Source: [2],[3],[10],[11],[12],[13],[14],[15]

  3. Guiding Nuclear Technology Parameters for Georgia Tech DEMO Design

  4. Stress Calculations • Both primary stress on coolant tubes and thermal breeding stress due to temperature gradient across the coolant tubes were calculated [4] • Primary stress was found to be 276.70 MPa and thermal bending stress is found to be 3.43 MPa • Combined stress is calculated as ~ 280.13 MPa • Standardized ASME stress is 270 MPa combined [1] • ~ 3.68% difference; can adjust by decreasing coolant tube radius

  5. Stress Calculations here, Pc = 2000 psi Pdis = 8033 psirc = 6 mm tc = 1.5 mm fpfc = 4/3 q’’= 0.35 Mw/m^2 α = 17*10^-6 1/K q’’’ = 5.2 MW/m^3 κ = 320 W/mK ν = 0.3 E = 128 GPa = (276.7 + 3.43) Mpa = 280.13 MPa

  6. First Wall Lifetime Radiation Damage here, Pfus=1500 MW R = 6.5 m ap=1.5 m 𝜅 = 2.5 = 1.64 MW/m^2 W Limit: 50 dpa corresponds to 10 MW-yr/m2 [6] (10 MW-yr/m2) / (1.64 MW/m2) = 6 years

  7. Thickness of Blanket and Shield An MCNP code was written for a simplified tokamak design in order to calculate the tritium breeding ratio and 1 MeV neutron flux at the magnets. The model included a 5mm tungsten first wall, a 0.5m Li4SiO4 blanket, a 0.5m mixture of B4C and stainless steel with a ratio of 1:1, and water cooling channels.

  8. Thickness of Blanket and Shield

  9. Thickness of Blanket and Shield Based on the results of the simulation the tritium breeding ratio was determined to be 1.16. Accounting for the volume of steel the TBR becomes 0.99 [4]. The fluxes of 1 MeV neutrons at the toroidal and poloidal magnets were 9.18E9 and 2.63E10 n/m^2s for the avg neutron wall load and 1.89E10 and 5.054E10 n/m^2s for the peak neutron wall load. This is well below the limit of 10^21 n/m^2 > 1 MeV in order for the magnets to carry the required current. [4]

  10. Peak Temperature In Blanket Number of Tubes =39000 [2] Number of sections of tokamak=9 Coolant = water @ 2000 psi; Inlet Temp=0 C Mass flow rate =8508 kg/s per section First Wall Heat Flux = 0.35 MW/m^2 Tube material =CuCrZr Breeder = Li4SiO4

  11. Peak Temperature of Blanket

  12. Peak Temperature of Blanket Based on the calculations shown the peak temperature of the blanket will be 830 degrees celsius which is below the melting temperature of the breeding material. [4] However; the temperature of the blanket will increase as the coolant nears the exit due to the increase in coolant temperature. At the exit the blanket temperature will be 1010 degrees celsius which is still below the melting temperature of the breeder but outside of the optimal operating temperature.

  13. Heat Removal in Divertor Divertor Heat Flux = 3.5 MW/m^2 Divertor dimensions r=6.5 m w=.025 m t=.005 m [3] First wall Material = Tungsten [3] Coolant = water @ 2000 psi; Inlet Temp=0 C Tube Material = CuCrZr [3] Number of Tubes per section = 500 [5] Mass Flow Rate = 356 kg/s per section

  14. Heat Removal In Divertor

  15. Heat Removal In Divertor The temperature of the FW of the divertor needs to be around 1000 degrees celsius in order to have the tensile strength to withstand the forces acting on it. Based on the temperature and stress calculations previously shown the first wall will be able to cope with stresses of the tokamak. [1]

  16. References • http://conductivity-app.org/ Material properties • Yong Liu, Hongguang Yang, Qin Zhan, Shanshan Liu, Bin Zhu, Yisheng Zhang, Measurements of the effective thermal conductivity of a non-compressed Li4SiO4 pebble bed, Fusion Engineering and Design, Volume 125, 2017, Pages 545-550, ISSN 0920-3796 • J.H. You, E. Visca, T. Barrett, B. Böswirth, F. Crescenzi, F. Domptail, M. Fursdon, F. Gallay, B-E. Ghidersa, H. Greuner, M. Li, A.v. Müller, J. Reiser, M. Richou, S. Roccella, Ch. Vorpahl, European divertor target concepts for DEMO: Design rationales and high heat flux performance, Nuclear Materials and Energy, Volume 16, 2018, Pages 1-11, ISSN 2352-1791 • Stacey, Weston M. Fusion: an Introduction to the Physics and Technology of Magnetic Confinement Fusion. Wiley-VCH, 2011. • L. A. El-Guebaly, ARIES Team, and FNSF Team. (2018) Nuclear Assessment to Support ARIES Power Plants and Next-Step Facilities: Emerging Challenges and Lessons Learned. Fusion Science and Technology 74:4, pages 340-369. • Tanno, Takashi et al., Precipitation of Solid Transmutation Elements in Irradiated Tungsten Alloys, Materials Transactions, Vol, 49, No. 10(2008) pp. 2259 to 2264, The Japan Institute of Metals 2008. • L. Ciupinski et al., TME observation of radiation damage in tungsten irradiated by 20 MeV W ions, 19th International Workshop on Inelastic Ion-Surface Collisions (IISC-19), Frauenchiemsee, Germany, 16-21 September, 2012. • A.F. Rowcliffe et al., Materials-engineering challenges for the fusion core and lifetime components of the fusion nuclear facility, Nuclear Materials and Energy, Volume 16, August 2018, Pages 82-87. • R. Mitteau et al., The design of the ITER first wall panels, Fusion Engineering and Design, Volume 88, Issues 6-8, October 2013, Pages 568 - 570. • G. Federici, C. Bachmann, W. Biel, L. Boccaccini, F. Cismondi, S. Ciattaglia, M. Coleman, C. Day, E. Diegele, T. Franke, M. Grattarola, H. Hurzlmeier, A. Ibarra, A. Loving, F. Maviglia, B. Meszaros, C. Morlock, M. Rieth, M. Shannon, N. Taylor, M.Q. Tran, J.H. You, R. Wenninger, L. Zani, Overview of the design approach and prioritization of R&D activities towards an EU DEMO, Fusion Engineering and Design, Volumes 109–111, Part B, 2016, Pages 1464-1474, ISSN 0920-3796 • K.M. Feng, G.S. Zhang, G.Y. Zheng, Z. Zhao, T. Yuan, Z.Q. Li, G.Z. Sheng, C.H. Pan, Conceptual design study of fusion DEMO plant at SWIP, Fusion Engineering and Design, Volume 84, Issue 12, 2009, Pages 2109-2113, ISSN 0920-3796 • Keeman Kim, Hyoung Chan Kim, Sangjun Oh, Young Seok Lee, Jun Ho Yeom, Kihak Im, Gyung-Su Lee, George Neilson, Charles Kessel, Thomas Brown, Peter Titus, A preliminary conceptual design study for Korean fusion DEMO reactor, Fusion Engineering and Design, Volume 88, Issues 6–8, 2013, Pages 488-491, ISSN 0920-3796 • Yu.A. Sokolov, Overview of the Russian Demo plant study, Fusion Engineering and Design, Volume 29, 1995, Pages 18-27, ISSN 0920-3796 • K. Tobita, S. Nishio, M. Enoeda, M. Sato, T. Isono, S. Sakurai, H. Nakamura, S. Sato, S. Suzuki, M. Ando, K. Ezato, T. Hayashi, T. Hayashi, T. Hirose, T. Inoue, Y. Kawamura, N. Koizumi, Y. Kudo, R. Kurihara, T. Kuroda, M. Matsukawa, K. Mouri, Y. Nakamura, M. Nishi, Y. Nomoto, J. Ohmori, N. Oyama, K. Sakamoto, T. Suzuki, M. Takechi, H. Tanigawa, K. Tsuchiya, D. Tsuru, Design study of fusion DEMO plant at JAERI, Fusion Engineering and Design, Volume 81, Issues 8–14, 2006, Pages 1151-1158 • L. Giancarli et al., Test Blanket modules in ITER: An overview on proposed designs and required DEMO-relevant materials, Journal of Nuclear Material Volumes 367-370, Part B, 1 August 2007, Pages 1271 - 1280.

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