Neutronics and Tritium Breeding Capability for Liquid Metal-Based Blanket (DCLL) L. El-Guebaly Fusion Technology Institute University of Wisconsin-Madison http://fti.neep.wisc.edu/UWNeutronicsCenterOfExcellence Contributors: S. Malang (Consultant), L. Waganer (Consultant), A. Rowcliffe (ORNL) FESAC TEC 2017 Panel Community Input Workshop Rockville, MD May 30 – June 1, 2017 Scope and Motivation • Transition from ITER to fusion DEMO and power plant requires development of tritium breeding blanket to generate tritium (T) in unprecedented large quantities to sustain the plasma operation. • Dual-coolant lead lithium (DCLL) blanket is prominent option for future US devices: tokamaks, spherical torii, and stellarators. • This talk : – Provides brief overview of breeding capability of lead lithium (PbLi) breeder under typical power plant operating conditions – Identifies uncertainties in tritium breeding prediction – Outlines critical and design variables that determine and control breeding level – Suggests developing several technologies to assure T self-sufficiency and mitigate risk of shortage or surplus of T – Mentions briefly related transforming technologies: – Tritium extraction – Additive manufacturing and nano-fabrication – Advanced ODS alloys. 2 Nuclear Assessment Involves Three Closely-Related Tasks Neutronics: Shielding: Activation: – Neutron wall loading distribution: peak – Shielding of permanent – Radioactive product inventory (for and average values components (for protection safety, environmental, and licensing assessments) – Tritium breeding ratio (TBR) for T-self against radiation) – Environmental impact of FNSF sufficiency – Radial and vertical build definition and radwaste classification – Radial and poloidal nuclear heating (for physics code, CAD drawings, – Decay heat (for thermal response distribution (for thermal hydraulic and systems code analysis) of components during LOCA/ analysis) – Neutron streaming through LOFA events) – Nuclear energy multiplication (for power assembly gaps and penetrations – Biological dose (for maintenance balance) (for protection of external components) crew, workers and public – Radiation damage to structural materials protection). (dpa, He production, H production) – Bioshield specifications. – Service lifetimes of all components based on neutron dose (for replacement frequency and cost). This talk focuses on T breeding issues – a subtask of neutronics assessment for fusion devices (such as power plants, DEMO, and FNSFs) seeking T self-sufficiency Refs: L. El-Guebaly, “Progressive Steps Towards Integral Nuclear Assessments for Fusion Devices,” Transactions of ANS winter meeting, Washington, DC, November 10-14, 2013. Laila El-Guebaly, “Overview of ARIES Nuclear Assessments: Neutronics, Shielding,3 and Activation,” Progress in Nuclear Science and Technology 4, 118-121 (2014). DOI: 10.15669. Tritium Resources and Economics • Cost of purchasing T ($30k to ~$118k per gram)1 is expensive enough to enable defining mission of DEMO and power plants (e.g., T self-sufficiency) and designing some components around TBR. • ITER will consume almost all T recovered from CANDU reactors (~1.7 kg/y). • Other sources of T exist in U.S. and abroad, but they are limited in supply, classified, uneconomical, and/or inaccessible for general use2. • For these reasons, fusion devices generating substantial fusion power and consuming 10s-100s kg of T annually must breed their own T in blanket (TBR > 1) to negate risk of relying on external supplies to provide/control essential fuel of machine. 1500 2000 MW P • TBR should be estimated with high fidelity as deficiencies in 1250 f $118k / g of T TBR prediction represent significant economic burden. 1000 • A small 1% error in TBR estimation is equivalent to ~1.1 kg of 750 T/FPY for 2000 MW fusion power. 500 250 • For T unit cost ranging from ~$30k to ~$118k per gram, 1% Tritium Cost ($M / FPY) $30k / g of T 0 deficiency implies an additional FPY operational cost of 0 5 10 $33-131M to purchase T from external sources. Error in TBR Estimation (%) Refs: 1- DOE Inspector General Report DOE IG-0632, December 2003; www.ig.energy.gov/documents/CalenderYear2003/ig-0632.pdf4 . 2- The JASON Report on Tritium. MITRE Corporation, JSR-11-345, November 2011. Variety of Liquid and Ceramic Breeders Proposed to Breed Tritium since 1970 2 84.3 Pb Li 15.7 84.3 (90% Li-6) Li Pb (90% Li-6) 15.7 75 Li Sn 1.5 O 2 25 Li Li Flibe (90% Li-6) Liquid Metals: PbLi, Li 4 3 3 SiO Flibe ZrO 4 TiO 2 2 2 Li Li 1 Li AlO Molten Salts: LiSn, Flibe, Flinabe Li 75 Sn Ceramic Breeders: Li2O, Li4SiO4, 25 Local Tritium Breeding Ratio 0.5 Li Li2ZrO3, Li2TiO3, LiAlO2 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Ceramic Breeders Liquid Breeders Tritium breeding ratio (TBR) is key metric for T self-sufficiency (TBR is ratio of T bred in blanket relative to T burned in plasma) (Local TBR for fairly thick cylindrical blanket (2 m), no structure, no multiplier, natural Li unless indicated, 100% dense materials at room temperature. Some breeders may contain neutron multiplier (e.g., Pb in LiPb and Be in Flibe). Ref: L. A. EL-GUEBALY, L. V. BOCCACCINI, R. J. KURTZ, and L. M. WAGANER, “Technology-Related Challenges Facing Fusion Power Plants,” Chapter in book: Fusion Energy and Power: Applications, Technologies and Challenges.5 NOVA Science Publishers, Inc.: Hauppauge, New York, USA. ISBN: 978-1-63482-579-5 (2015). PbLi* – Most Popular Liquid Metal Worldwide • In US, PbLi is preferred breeder for future devices: Fusion Nuclear Science Facilities (FNSF) and power plants based on tokamak, spherical torus, and stellarator concepts. • In Europe, EUROfusion is currently examining dual coolant lead lithium (DCLL), water-cooled lead lithium (WCLL), and He-cooled lead lithium (HCLL) blankets for EU DEMO and/or power plants. • Japan prefers molten salts (Flibe/Flinabe) with some interest in PbLi blankets. • China is developing both HCLL and DCLL blankets for fusion applications. • Most PbLi-based blankets are supplemented with water or helium coolant, as in WCLL, HCLL, and DCLL. * Refs: S. Malang et al. (2011). “Development of the lead lithium (DCLL) blanket concept,’’ Fusion Science and Technology, 60, 249. L. A. EL-GUEBALY, L. V. BOCCACCINI, R. J. KURTZ, and L. M. WAGANER, “Technology-Related Challenges Facing Fusion Power Plants,” Chapter in book: Fusion Energy and Power: Applications, Technologies and Challenges. NOVA Science Publishers, Inc.: Hauppauge, New York, USA. ISBN: 978-1-63482-579-5 (2015). Laila A. El-Guebaly, “Fifty Years of Magnetic Fusion Research (1958-2008):6 Brief Historical Overview and Discussion of Future Trends.” Energies 2010, 3 (6), 1067-1086 (June 2010). Available at: http://www.mdpi.com/1996-1073/3/6/ PbLi: Most Popular Liquid Metal Worldwide (Cont.) (US Tokamaks) PbLi Li CB ARIES-I ARIES-ACT-2 1000 MWe, 6.75 m R, 4.5 A 1000 MWe, 9.75 m R, 4 A 1.9% βT, 21 T Bc, 16 TFC 1.5% βT, 14.4 T Bc, 16 TFC SiC/Li2ZrO3/He/Be Blanket FS/LiPb/He Blanket 49% ηth, 76% Avail 44% ηth, 85% Avail 87 mills/kWh ~91 mills/kWh ARIES-II ARIES-IV STARFIRE 1000 MWe, 5.6 m R, 4 A 1000 MWe, 6 m R, 4 A 1st Steady-State Design 3.4% βT, 16 T Bc, 16 TFC 3.4% βT, 16 T Bc, 16 TFC 1200 MWe, 7 m R, 3.6 A V/Li Blanket SiC/Li2O/He/Be Blanket 6.7% βT, 11 T Bc, 12 TFC 46% ηth, 76% Avail 49% ηth, 76% Avail PCS/LiAlO2/H2O/Be Blanket 76 mills/kWh 68 mills/kWh 36% ηth, 75% Avail Tokamak (29) 110 mills/kWh 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 13 ARIES-ACT Aggressive and Conservative Tokamaks (UCSD) DEMO-S steady state DEMO Russia SlimCS Compact low-A DEMO Japan FDS-II China power plant China VECTOR VEry Compact TOkamak Reactor Japan ARIES-AT DEMO2001 Japan 1000 MW , 5.5 m R, 4 A PPCS Conceptual Study of Fusion Power Plants EU e ARIES-AT Advanced Tokamak (UCSD) 9.2% βT, 11 T Bc, 16 TFC-HT A-SSTR2 Combine advantages of A-SSTR and DREAM Japan SiC/LiPb Blanket ARIES-RS Reversed-Shear tokamak (UCSD) w/ YBCO 59% ηth, 85% Avail A-SSTR Advanced Steady State Tokamak Japan HTS Magnet 1970s UWMAK series DREAM Drastically Easy Maintenance Tokamak Japan 48 mills/kWh contributed to basic understanding CREST Compact Reversed Shear Tokamak Japan of fusion power plant design and PULSAR-I/II pulsed tokamak (UCSD) ARIES-IV Second-stability tokamak (UCLA) technology and uncovered ARIES-II Second-stability tokamak (UCLA) undesirable aspects of: ARIES-III D-3He-fuelled tokamak (UCLA) SSTR steady state tokamak Japan - Pulsed operation ARIES-I First-stability tokamak (UCLA) - Low power density machine Apollo D-3He Fuelled Tokamak (UW) Wildcat catalyzed D-D tokamak (ANL) - Plasma impurity control problems STARFIRE Commercial Tokamak Fusion Power Plant (ANL) - Maintainability issues. NUWMAK University of Wisconsin Tokamak (UW) Russia TVE-2500 high temperature power plant with direct conversion ARIES-RS Many proposed technologies are still UWMAK-III University of Wisconsin Tokamak (UW) 1000 MWe, 5.5 m R, 4 A considered in recent designs: 316-SS, Li UWMAK-II University of Wisconsin Tokamak (UW) 5% βT, 16 T Bc, 16 TFC,V/Li and LiPb breeders, solid breeders, Be A Fusion Power Plant (PPPL) UWMAK-I University of Wisconsin Tokamak (UW) 46% ηth, 76% Avail multiplier, NbTi and Nb3Sn S/C, solid Premak University of Wisconsin Tokamak (UW) 7 76 mills/kWh and liquid Li divertors. 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 04 05 06 07 08 09 10 11 12 13 calendar year PbLi: Most Popular Liquid Metal Worldwide (Cont.) (International