Activities of Helmholtz Association research centers on fast reactors

Activities of Helmholtz Association research centers on fast reactors A. Rineiski, KIT, Karlsruhe, Germany 50 th IAEA TWG-FR Meeting, Vienna, May, 2017 INSTITUTE FOR NUCLEAR AND ENERGY TECHNOLOGIES KIT The Research University in the Helmholtz Association www.kit.edu

Nuclear in Helmholtz Association (HGF) Helmholtz Association about 30.000 employees; about 3 billion yearly budget Nuclear Research at HGF: at Jülich, Dresden Rossendorf, Karlsruhe about 300 scientists, about 50% at Karlsruhe Research at HGF related to fast reactors (FRs) Helmholtz Center Dresden Rossendorf (HZDR) Karlsruher Institute of Technology (KIT) 2

KIT Education and Training Frédéric Joliot / Otto Hahn Summer School AREVA Nuclear Professional School (ANPS) KIT Batchelor & Master Courses Training within 7th FP of EU Support of an IAEA training course at ICTP Training within working groups 3

General Situation for HGF Research Institutes POF-III program of HGF established for 2015-2019, main nuclear topics: Safety research with a focus on LWR severe accidents Research on nuclear waste disposal including transmutation FR safety studies Advanced fuel cycle studies for P&T Nuclear phase out in Germany by 2022, more attention to decommissioning, intermediate storage, final disposal Reorientation of FR-related experimental program to other R&D fields, e.g. liquid metal technology for alternative energy options Possible extension of POF-III under discussion Future planning on Nuclear Safety Research for 2020+ (POF-IV): Decommissioning and waste management Safety studies to be continued, in particular as contribution to EU and International activities, including FR safety studies Studies on advanced fuel cycle studies in the future: under discussion 4

In order to identify potential advantages that P&T may offer for waste management strategy in Germany, the Federal Ministry of Economics and Technology (BMWi) and the Federal Ministry of Education and Research (BMBF) launched and granted in the period 2012-2014 an interdisciplinary research project, to support the decision on merit of P&T implementation. Several scenarios in a nuclear phase-out context have been investigated: Scenario 1: no actions in support of P&T strategy Scenario 2: only R&D activities related to P&T (i.e. postponing the decision of P&T implementation) Scenario 3: Isolated application of P&T in a phase-out context: Pu and MA burner Scenario 4: Implementation of P&T in a regional (e.g. European) context: MA burner 2 main options considered for Scenaruo s 3 and 4: TRU in U-free matrix (MgO, Mo, ) ADS EFIT-like (400 MWth) TRU in U matrix Low CR FRs ASTRID-like with reduced core height and power (1200 MWth): to allow higher TRU content, but negative void coefficient Additional options: MSFR (not studied in detail): Pu and MA burner, ESFR-like with reduced core by 20% core height (2880 MWth): MA burner Similar performance per unit: larger reactors allow smaller amounts of MAs 5 The German P&T study (2012-2014)

European Sustainable Nuclear Industrial Initiative (ESNII) systems and MSFR: main reactor options considered SFR, ADS/LFR, GFR: near term MSFR: long term 6

7 EU projects on Fast Reactors SFR: ASTRID of ESNII+ (ends in 2017): reactor physics, transient analyses, severe accidents: SAS+SIMMER and SIMMER-only routes og calculations, workshop on FR severe accident simulations at KIT Core melting postponed after ULOF in case of low void cores, but not excluded under conservative assumptions. ESFR-SMART (starts in 2017): core optimization, reactor physics, transients, severe acidents and mechanical work potential, safety-related experiments and simulations, workshops on safety studies GFR/LFR: GoFastR/ALLEGRO => parst of ESNII+ MYRRHA: SEARCH, MAXSIMA => MYRTE: experiments on blockage at KIT (KALLA facility), simulation of ADS reactor physics experiments SAMOFAR : Molten Salt Fast Reactor fuel draining into a special tank, decay heat removal from the tank PELGRIMM : Sphere-Pac Fuels for SFRs: safety studies at KIT (finished) THINS : Thermal-hydraulics Advanced Reactors=>SESAME JASMIN: SFR safety and extension of the ASTEC code SFRs (ASTEC-Na)

FR Activities with EU/International partners CEA: ARDECo cooperation on ASTRID Reactor safety, SIMMER code developement (KIT) Eddy Current Flow Meter development, sodium flow & magnetic field instrumentation for a Na facility (HZDR) IAEA: EBR-II (FR17 papers), new proposals on CEFR (KIT, HZDR, GRS) Sodium Properties, Source Term in SFR ADS physics: simulation of experiments NEA: Advanced fuel cycle scenario s for nuclear waste management C/E analyses on on SFR MA-bearing fuel pin behavior Safety parameters for SFR and MSR and their uncertainties Other bilateral cooperations: SFR/LFR safety KIT-NRA study on thermal to mechanical conversion ratio: FR17 paper 8

Measurement technologies for liquid metals (also non-nuclear) (a) Flow rate electro-magnetic, UTT, momentum based.. Visualization techniques direct X-Ray tomography indirect CIFT, Utra-sound-transient time (UTT),. Velocity direct Pitot-Tube (Dp) magnetic potential probes (MPP) fibre-mechanics Thermocouples 5mm 2 Thermocouples 0.2mm pressure orifice probe tip magnet tube sup. tube probe shaft match Non-intrusive Ultra-sound doppler velocimetry (UDV), multi units mapping Surfaces /2-phase direct resistance probes indirect X-ray, UTT optic means for surfaces Neutronic core monitoring fission chambers semi-conductors- SiC based- diode (SPND) (neutron-generator available ) SPND 9

Infrastructures (also non-nuclear) loop facilities Table-top/proof of principle GALINKA Laboratory scale Pilot-scale (proto-typical) THEADES TELEMAT GaInSn ALINA 44t, 1MW, 6bar 450 C PbBi PbBi KASOLA 7t, 2MW, 6 bar, 550 C Available liquids Lead, PbBi-eutectics, tin Gallium-Indium, tin Sodium, NaK, THESYS 2t, ~200kW, 4 bar, 450 C 10

LM experimental facilities at KIT ALINA SOMMER THESYS THEADES Thermal-hydraulics + MHD + Processes (IKET) KASOLA Thermal-hydraulics + Components (INR) ATEFA DITEFA GALINKA(x2) HELIS MEKKA LBE + Pb Na + NaK COSTA(x8) GESA FRETHME Materials (IHM) CORELLA SOLTEC(x3) GaInSn Sn, PbLi PICOLO Corrosion CORRIDA (IAM-WPT) KOSIMA TELEMAT CRISLA 11

19-rod bundle with wire spacers D=8.2 mm LBE FLOW d=2.2 mm LBE FLOW H=328 mm 12 LBE FLOW P=10.49 mm

Heat transfer coefficient: Nu vs. Pe Flow development effects Low Nu: best agreement with Kazimi Hot at center, cold at edge +34.6% +5.2% 13

Local flow blockages (small) LBE 1 7 C1 E1 Steel casing (SLM) filled with low thermal conductivity material 14

Local flow blockages (small) 1 7 DT=ca. 200 K for full power (1MWm 2 ) Q Global effect in pressure drop is negligible cannot be detected Local temperature increase is significant, still acceptable 15

Local flow blockages (large) LBE 7L 7L L=H/6=54.6mm L C6 Experimental campaign completed in May 2017 Constructed in 3 parts around rod #1, no contact with flowing LBE Detailed evaluation ongoing DT ~ 200 K at ~25% power 16

Blockage formation by particles 19-pin blockage formation rod bundle in ware for optical access to the sub channels. Study characteristics of blockages Location (center, edge) Size (how many sub-channels) Porosity (from 0% to 100%) Length (relative to wire pitch H) Materials (density) Identify a realistic worst-case scenario for blockages Blockage formation experiment 17

LBE Inter-wrapper flow experiment IWF occurs in NC-DHR (Kamide et al) IWF can reduce hot spots (e.g. in blocked FAs) 3x7 heated rods and inter-assemblies gap (3.0 mm) 3x independent power supplies for asymmetric tests LBE Reference case is MYRRHA (LBE) 18

KIT SIMMER Application MAKSIMA EU-Project MAXSIMA Severe Accidents in MYRRHA Reactor (forerunner projects: XT-ADS, CDT, SEARCH) Study of transient behavior for TIB initiator Subchannel geometrical arrangement Momentum exchange model developed for the Cross Flow Results have been confirmed by a subchannel code Steady State Case Points : SIMMER : Subchanflow TIB Case 19

3D Models: blockage studies in MYRRHA with intra-sa gaps The coolant flow through inter-wrapper gaps between subassemblies can retard or even prevent the canwall failure and its propagation. Gaps were modeled with special meshes in a 2D case providing quite different results as compared with implicit option. The gaps are explicitly modeled also in 3D which takes account of 3-D heterogeneity and needs more but acceptable CPU time and computer memory. S a a K b=s/(2a) a a I K (r) (r) b=s/(2a) (r) (r) (r) (r) a a (r) (r) I 20 with left, right and back CWs, (r) with left and right CWs with left, right and front CWs, (Others) filled with 100% coolant (hexcan gaps)

DYN3D Developement (IP) for FRs at HZDR 13 4 Coupled codes developed at HZDR for 15 16 2 safety analysis of LWR cores Validation studies done for for PWR, BWR, VVER reactors 9 10 12 11 1 5 6 8 7 3 Extension of DYN3D to FRs Modeling for initiation transient phase (before fuel melting) XS generation methodology Validation of DYN3D by fast reactor experiments models for thermal expansion effects Axial expansion of the fuel rods Radial expansion of the diagrid DYN3D 3D reactor dynamics (steady-state, transient) cartesian & hexagonal geometry Neutron kinetics neutron flux distribution for power calculation Thermal hydraulics coolant parameters distribution for heat removal and feedback 14 21

The reactor dynamics code DYN3D DYN3D 3D Reactor dynamics code for steady-state and transient calculations in Cartesian and hexagonal geometry Neutron kinetics Thermal hydraulics Neutron flux distribution for power calculation Close Interaction Coolant parameters distribution for heat removal and feedback Developed at HZDR for the safety analysis of light water reactors cores Code versions for Western Pressurized Water (PWR) and Boiling Water Reactors (BWR) and Russian VVER type reactors PWR BWR VVER 22

SCRAM Temperature ( C) Radial expansion model for the diagrid Assumption Driven by the average inlet coolant temperature Sub-assembly pitch size expands uniformly (unlike axial expansion with SA-wise approach and XS mixing) Model for DYN3D Adjusting pitch size at each time step XS parametrized with relative diagrid expansion 1D heat structure model for the time delay 410 400 390 380 370 360 Sodium Diagrid 350 0 100 200 300 400 500 Time (s) Radial expansion coeff. dl/l T (K) Radial expansion reactivity worth (pcm) (pcm/k) OECD/NEA large oxide core ref. 668 Serpent DYN3D Difference Serpent DYN3D 1.00% 1153-429 ± 4-408 21-0.885-0.841 Phenix EOL core ref. 523 Serpent DYN3D Difference Serpent DYN3D 1.00% 1032-529 ± 4-507 22-1.039-0.996 23

Initialization SAS-SFR PARCS SAS-SFR development (IP) at KIT Modeling for the initiation transient phase (IP):before massive material melting Models for MOX fuel pin behavior and 1D fuel- and steel material relocation embedded in the code Coupling of SAS-SFR (based on SAS4A) with a XS-generation model and PARCS neutron flux calculation model Development is being tested for ESFR WH (large size SFR) Python interface: data processing and visualization PARCS_SS PARCS_TS SA power, MW SAS-SFR_SS SAS-SFR_TS Sub-assemblies of inner core (225 SAs) Sub-assemblies of outer core (228 SAs) Control rods (24 CSDs) Shutdown rods (9 DSDs) Radial steel reflector sub-assemblies and central dummy channel 24

SIMMER development at KIT Code developement together with JAEA, CEA and other partners Coupling to external structure mechanics and fuel behavior models channel approach for IP with a SIMMER model per SA, to faciliate coupling for TP (together with CEA), metal fuel as results of EBR-II CRP (together with Kyushu University, KU) subchannel models for blockage, new feedbacks (core and Control Rod driveline thermal expansion,heterogenity effect variation due to melting,...) coupling to structure mechanics for expansion phase simulation Fluid Dynamics (TH) 8+ velocity fields (7+ for liquid, 1 for gas) Multi-phase, multi-component flow Phase transitions Flow regime (pool-channel) Interfacial area tracking EOS (various fuels, coolants, gases) Heat and mass and momentum transfer C 4 P 1968/560 Group Master Library Basis: JEFF, JENDL, ENDF/B Full Range Neutron Spectrum Structure model General structure model Pin model (new fuels-development) Loop model (IHX & pumps) Axial + radial heat transfer Virtual structure model Structure disintegration Freezing on structures Neutronics Neutron transport (diffusion) improved quasi-static method Cross-section generation Heterogeneity treatment Decay heating movable neutron precursors external n.-source & source importance PARTISN neutron transport solver, heterogeneity, thermal exp. feedbacks 25

SIMMER application to EBR-II IAEA CRP on EBR-II shut-down heat removal tests extended SIMMER has been validated by comparison with experimental data. Shielding zone (excluded from calculation) IHX Core Na plenum Z-pipe Pum p Blanket Low pressure inlet pipe High pressure inlet pipe 26

Core Thermal Expansion feedback reactivity - 4 Z-Pipe inlet XX09 (core top) Global effect 27 Local effect

Accidental safety analysis- Containment behavior Sodium fires (physical & chemical processes) in severe accidents Approach: Validation of CONTAIN LMR for single effect phenomena atmosphere thermodynamics including condensation and vaporization of sodium; reactions between sodium and oxygen or water (sodium spray or pool fires); sodium aerosol behavior. Reference FAUNA experiments @ KIT CONTAIN- Features oxygen diffusion model (pool-fire), spray fire models sodium combustion chemistry Height aerosol description (diffusion, graviational settling, agglomeration) experiment No. F1 F2 F3 pool surface (m 2 ) 2 2 12 sodium (kg) 150 250 500 pool depth (mm) 90 150 50 c O2 (vol.% ) 19-22 17-25 15-25 T start ( C) 550 550 550 T Pan ( C) 250 250 250 More information: Gordeev et al. 2014, CONTAIN-LMR simulation of sodium aerosol behaviour during sodium fires, ICAPP 2014. Cherdron W, 1985. Thermodynamic Consequences of Sodium Spray Fires in Closed Containments. Part 1. KfK 3829, Juni 1985 28 Burn chamber: Volume 220m 3 Diameter 6m 6m

Reactor data management KIT developed an approach and a format for reactor data specifications in 2015-2016: details in the KIT presentation at the meeting on FR simulators early 2016 This approach is potentially useful for future IAEA CRPs KIT is in contact with IAEA TWG-FR, considerations to proceed in 2017 with previous CRP models (BN-600, Phenix EOL, EBR-II): to establish data and interfaces to codes The outcomes: to discuss at the next TWG-FR meeting in 2018 29

Concluding remarks FR studies in Germany: in view of FR transmutation potential Main effort on safety studies for EU and International projects/partners Also effort on possible transmutation-related fuel cycles Safety: main effort on SFR, MYRRHA, other ESNII systems FR safety studies to be continued also for the next term (after 2020) Fuel cycle and transmutation studies run for the current POF term, to be discussed for the future FR experimental facilities: existing and new ones supported by non-nuclear activities (concentrated solar energy, etc.) Effort on safety code developement and application Education and training 30