INL experimental and analytical capabilities for the DCLL
INL experimental and analytical capabilities for the DCLL concept www.inl.gov 2nd EU-US DCLL Workshop Brad J. Merrill, Fusion Safety Program November 14-15th, 2014 University of California, Los Angeles Presentation Outline 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 2 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 INL-Site INL-Site
Idaho Cleanup Project Naval Reactors Facility the National Nuclear Laboratory 3 FSPs STAR Lab located at the INL Site 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, T2) typically specified as either: Contact dose (< 1.5 mSv/hr at 30 cm, normally < 100 Sv/hr), Annual worker dose (< 7 mSv/yr) Total inventory (1.5 g T2 or 5.6x105 GBq) STAR 4 Tritium Solubility testing in Lead-Lithium Eutectic (LLE) under US-JA TITAN Collaboration
5 6 Tritium Gas Absorption Permeation (TGAP) Experiment Unique capabilities 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 7 Blank Tests Results - Qualify System w/o LLE Tritium Concentration (Ci/m3) 100 10 10-2 0.01 Pa 0.15 Pa 0.001 Pa 10-3 10-4 10-5
TMAP model of TGAP 2.4 Pa -1 TMAP - Symbols 0 1000 2000 3000 4000 5000 Time (s) Permeation results with empty test section for -Fe membrane can be matched well with TMAPs surface kinetics model (requires a multiplier of 0.1 on surface coefficients kd & kr) 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 8 Role of H2 at Surface of Membrane in TLLE
Secondary H (100 Pa) changes T Results for Blank Tests 2 permeation from surface to diffusion limited permeation & reduces membrane T concentration Hydrogen No hydrogen H2 H 2.4 Pa HT 0.15 Pa T2 H2 HT T T2 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) 9 Initial LLE Transport Properties Determined by Analysis Tritium permeation through 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) (1 mm) -Fe + (6 mm) LLE 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 10 TGAP Tritium Permeation Campaign for Tungsten 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 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 MELCOR: Integrated Physics and Coupled During post ITER EDA , a multi-fluid version of MELCOR 1.8.5 was developed to assess the safety of Advanced Phenomena Fusion Reactor Concepts. This version of MELCOR uses the ATHENA/RELAP fluid Fusion ITER Modifications property subroutines which include 10 fluids EOS modifications for water freezing & ice layer SPARC FliBe, Hydrogen, Helium, Lithium, NaK, Nitrogen, formation Potassium, PbLi, Sodium, Water BWRSAR CORSOR C, Be, W oxidation (INL correlations) Flow choking changed to Fauske model Aerosol transport module MELCOR modifications for gas
Lithium pool fire model similar to LITFIRE mixtures, turbulentCORCON & inertial deposition Modifications to Multi-fluids MELCOR 1.8.5 for NGNP Enclosure thermal radiation heat transport MAEROS application were made Flow Boiling heat transfer HECTR VANESA Gaseous binary diffusion was added to the mass and HTO transport In 2008, the NRC requested that all model energy conservation equations Used in CONTAIN ITER Safety Documents fusion modifications be included in Graphite oxidation and combustion models 2.x F95 version the MELCOR
TRENDS NSSR 1&2, GSSR, RPrS (after being pedigreed Dust re-suspension version available. A and placedMELPROG in INL Software QA system asPreliminary a QL1 Used in advanced reactor design studies: (Li/ paperAPEX at the Evolve 22nd International safety code) VICTORIA W), ARIES-AT & CS (Dual Coolant Lead Lithium Conference onDCLL), Nuclear Engineering US ITERIFCI DCLL Test Blanket Module on (TBM) sodium modeling by SNL-NM DEBRIS (2014) Pool F.P. Scrubbing BWR Plant Code
1.8.2 1.8.3 1.8.4 1.8.5 1.8.6 (2005) 2.x (2008) 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 Internal PbLi flow All ferritic steel
structures are He-cooled at 8 MPa, 350-410C PbLi self-cooled flows in poloidal direction at 2 MPa, 360-470C He out He in PbLi concentric inlet/outlet pipe 14 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. 102 101.0 101 Mass released (g) Pressure (kPa)
1000 2000 Time (s) 3000 15 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. 0.10 Fraction mobilized Temperature (C) 800 600 PbLi pool 400 200 Port plug 0 Inter-space wall
30 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: FSPs 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 existing Laboratory collaborations to leverage other The NationalinNuclear institution's capabilities and reduce duplication of effort. STAR will move towards being more effective FES User Facility. Our website is at: https://inlportal.inl.gov/portal/server.pt/community/fusion_safety 17
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