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Www.inl.gov INL experimental and analytical capabilities for the DCLL concept 2 nd EU-US DCLL Workshop Brad J. Merrill, Fusion Safety Program November.

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Presentation on theme: "Www.inl.gov INL experimental and analytical capabilities for the DCLL concept 2 nd EU-US DCLL Workshop Brad J. Merrill, Fusion Safety Program November."— Presentation transcript:

1 www.inl.gov INL experimental and analytical capabilities for the DCLL concept 2 nd EU-US DCLL Workshop Brad J. Merrill, Fusion Safety Program November 14-15 th, 2014 University of California, Los Angeles

2 2 Describe ongoing FSP experimental capabilities in tritium and materials safety research (in particular PbLi) through international collaborations Discuss computer codes developed by the FSP for accident analyses Presentation Outline

3 Idaho National Laboratory… 890 square miles 111 miles of electrical transmission and distribution lines 579 buildings 177 miles of paved roads 14 miles of railroad lines …the National Nuclear Laboratory Naval Reactors Facility Idaho Cleanup Project INL-Site 3

4 The Safety and Tritium Applied Research (STAR) facility is an Office of Science facility commissioned in November 2001 STAR Facility – DOE less-than Hazard Category 3 Radiological Facility – Allows handling of radioisotopes at low radiation levels (e.g., W, Ni, Mo, T 2 ) typically specified as either:  Contact dose (< 1.5 mSv/hr at 30 cm, normally < 100 µSv/hr),  Annual worker dose (< 7 mSv/yr)  Total inventory (1.5 g T 2 or 5.6x10 5 GBq) 4 STAR FSP’s STAR Lab located at the INL Site…

5 5 Tritium Solubility testing in Lead-Lithium Eutectic (LLE) under US-JA TITAN Collaboration

6 6

7 Designed to measure transport properties (e.g. diffusivity, solubility, and permeability) of tritium at realistic blanket conditions (e.g. low tritium partial pressure < 100 Pa) Using tritium should lead to accurate measurements since it is easily detected Uniform temperature (+/- 10 C) within the test section utilizing 12” tube furnace Unique capabilities 7 Tritium Gas Absorption Permeation (TGAP) Experiment

8 8 Blank Tests Results - Qualify System w/o LLE TMAP model of TGAP Permeation results with empty test section for α-Fe membrane can be matched well with TMAP’s surface kinetics model (requires a multiplier of 0.1 on surface coefficients k d & k r ) Simple analytical expressions for permeation flux do not accurately capture the response of the entire system Low pressure (0.001 Pa) points are problematic – lower limits of experimental resolution The background concentration of hydrogen must be quantifiable 0 10002000300040005000 Time (s) 10 0 Tritium Concentration (Ci/m 3 ) 10 -1 10 -2 10 -3 10 -4 10 -5 2.4 Pa 0.15 Pa 0.001 Pa TMAP - Symbols 0.01 Pa

9 9 T2T2 H2H2 T H T2T2 HTHT H2H2 HTHT 0.15 Pa Hydrogen No hydrogen 2.4 Pa Secondary H 2 (100 Pa) changes T permeation from surface to diffusion limited permeation & reduces membrane T concentration Primary H surface concentration > 40 x T, steepens T gradient (speeds equilibrium in membrane) Secondary side H carries T off surface back into secondary (reduces T flux) Role of H 2 at Surface of Membrane in TLLE Results for Blank Tests

10 Tritium permeation through (1 mm) α-Fe + (6 mm) LLE 10 Due to the very low tritium pressures, moderate sweep gas flow rates and “comparatively” large TGAP test volumes, data analysis must be accomplished though computer code matching (Alice pointed out yesterday that this experiment tests liquid metal/solid metal interface assumption) Modeling results (Preliminary analysis) Tritium diffusivity in LLE: A factor of 2-3 higher value needed to fit exp. data than found in literature Tritium solubility in LLE: Similar to literature data (?) TGAP is being reconfigured so that the background hydrogen concentration will be known during experiments Initial LLE Transport Properties Determined by Analysis

11 Permeation test section developed under the Japan/US PHENIX collaboration by Shizuoka University capable of testing 6 mm diameter tungsten disks up to temperatures of 1000 C and low tritium partial pressures (< 100 Pa) With similar test sections under the NFRI-UCLA-INL collaboration, TGAP will be used to study tritium permeation through RAFM at low pressures and release from functional materials 11 TGAP Tritium Permeation Campaign for Tungsten

12 FSP Safety Codes Used in ITER and US DCLL TBM Accident Analyses MELCOR for fusion - a fully integrated, engineering level thermal- hydraulics computer code that models the progression of accidents in fission and now fusion power plants, including a spectrum of accident phenomena such as reactor cooling system and containment fluid flow, heat transfer, and aerosol transport (various fluids, including PbLi, can be modeled), ATHENA/RELAP – a multi-fluids thermal-hydraulics code developed for design and accident analysis of cooling systems fusion reactor systems, TMAP - a tritium migration code that treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by air flow within a given facility. 12

13 MELCOR: Integrated Physics and Coupled Phenomena 1.8.01.8.2 1.8.31.8.5 1.8.4 MELCOR Versions 1.8.11.7.1 MACCS1.8.6 (2005) 2.x (2008) Fusion ITER Modifications EOS modifications for water freezing & ice layer formation C, Be, W oxidation (INL correlations) Aerosol transport module modifications for gas mixtures, turbulent & inertial deposition Enclosure thermal radiation heat transport Flow Boiling heat transfer HTO transport model Used in ITER Safety Documents NSSR 1&2, GSSR, RPrS (after being pedigreed and placed in INL Software QA system as a QL1 safety code) Fusion ITER Modifications EOS modifications for water freezing & ice layer formation C, Be, W oxidation (INL correlations) Aerosol transport module modifications for gas mixtures, turbulent & inertial deposition Enclosure thermal radiation heat transport Flow Boiling heat transfer HTO transport model Used in ITER Safety Documents NSSR 1&2, GSSR, RPrS (after being pedigreed and placed in INL Software QA system as a QL1 safety code) During post ITER EDA, a multi-fluid version of MELCOR 1.8.5 was developed to assess the safety of Advanced Fusion Reactor Concepts. This version of MELCOR uses the ATHENA/RELAP fluid property subroutines which include 10 fluids – FliBe, Hydrogen, Helium, Lithium, NaK, Nitrogen, Potassium, PbLi, Sodium, Water – Flow choking changed to Fauske model – Lithium pool fire model similar to LITFIRE Modifications to Multi-fluids MELCOR 1.8.5 for NGNP application were made – Gaseous binary diffusion was added to the mass and energy conservation equations – Graphite oxidation and combustion models – Dust re-suspension Used in advanced reactor design studies: APEX Evolve (Li/W), ARIES-AT & CS (Dual Coolant Lead Lithium – DCLL), US ITER DCLL Test Blanket Module (TBM) During post ITER EDA, a multi-fluid version of MELCOR 1.8.5 was developed to assess the safety of Advanced Fusion Reactor Concepts. This version of MELCOR uses the ATHENA/RELAP fluid property subroutines which include 10 fluids – FliBe, Hydrogen, Helium, Lithium, NaK, Nitrogen, Potassium, PbLi, Sodium, Water – Flow choking changed to Fauske model – Lithium pool fire model similar to LITFIRE Modifications to Multi-fluids MELCOR 1.8.5 for NGNP application were made – Gaseous binary diffusion was added to the mass and energy conservation equations – Graphite oxidation and combustion models – Dust re-suspension Used in advanced reactor design studies: APEX Evolve (Li/W), ARIES-AT & CS (Dual Coolant Lead Lithium – DCLL), US ITER DCLL Test Blanket Module (TBM) In 2008, the NRC requested that all fusion modifications be included in the MELCOR 2.x F95 version Preliminary version available. A paper at the 22 nd International Conference on Nuclear Engineering on sodium modeling by SNL-NM (2014) In 2008, the NRC requested that all fusion modifications be included in the MELCOR 2.x F95 version Preliminary version available. A paper at the 22 nd International Conference on Nuclear Engineering on sodium modeling by SNL-NM (2014)

14 MELCOR Code Applied to US Test Blanket Module (TBM) Safety Assessment for ITER Evaluate consequences to ITER from accidents in the proposed US dual coolant lithium lead (DCLL) TBM (recently for EU HCLL) To date several accident scenarios have been investigated: – In-vessel TBM coolant leaks – In-TBM breeding zone coolant leaks – Ex-vessel TBM cooling system LOCA, LOFA, LOHS No significant impacts on ITER safety have been identified; a preliminary safety report has been published, INL/EXT-10-18169, and transmitted to ITER IO for TBM concept licensing All ferritic steel structures are He-cooled at 8 MPa, 350-410°C PbLi self-cooled flows in poloidal direction at 2 MPa, 360-470°C He out He in Internal PbLi flow PbLi concentric inlet/outlet pipe 14

15 15 Helium Leak into Inter-space Pressure/Release Results Gallery overpressure reaches ~ 700 Pa Of ~20.6 g of dust, ~0.4 g of tritium as HTO, and ~0.03 g of ITER activated corrosion products (ACP) transported into the gallery 11.7 g of dust, 0.03 g of tritium, as HTO, and 0.01 g of ACP are predicted to be released to the environment. 0100200300400500 Time (s) 99.5 100.0 100.5 101.0 Pressure (kPa) Gallery Cryostat room 0100020003000 Time (s) 10 -3 10 -2 10 -1 10 0 10 1 10 2 Mass released (g) Dust ACP HTO

16 PbLi Leak into Inter-space Pool Temperature/Mobilization Results Pool surface freezing in 23 hr, entire pool freezes by 130 hr, and by 250 hr the temperature drops to 110 C. Pb-210 and Hg-203 are mobilized (< 3% 1.8 Ci of the Pb-210 and < 10% of the 36 Ci of Hg-203) by diffusion of these isotopes in the pool and release from the surface by evaporation. Once the PbLi freezes, this diffusion and evaporation process should drop dramatically. 0100200300400 Time (hr) 0 200 400 600 800 Temperature (C) PbLi pool Bioshield Port plug Inter-space wall 0102030 Time (hr) 0.00 0.02 0.04 0.06 0.08 0.10 Fraction mobilized Po - 210 Hg - 203

17 Future Directions in FSE Research Strategy based on FES guidance and 2013 FES Peer Review Comments – Materials Research: Fusion materials, including tungsten irradiated, will be studied at high temperature and heat flux to measure tritium retention and permeation. Dust explosion measurements for fusion materials will continue in support of licensing and computer code development activities. New material diagnostics. – Code Development: for the near term, a newer version of MELCOR for ITER will to be completed that includes tritium transport and dust explosion models. Long- term: Multi-dimensional safety code capabilities needs to be developed that take advantage of parallel computing (example RELAP 7) – Risk and Licensing: FSP’s evolving failure rate database will be expanded to include maintenance data from existing tokamaks. Risk-informed safety analysis methods (example RISMC Toolkit) will be studied for application to an FNSF. Continue ASME codes and standards and licensing framework development. – Collaborations: Participation in existing collaborations to leverage other institution's capabilities and reduce duplication of effort. STAR will move towards being more effective FES User Facility. 17 The National Nuclear Laboratory Our website is at: https://inlportal.inl.gov/portal/server.pt/community/fusion_safety


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