Coupled Neutronic Fluid Dynamic Modelling of a Very High Temperature Reactor using FETCH Brendan Tollit KNOO PhD Student (BNFL/NEXIA Solutions funded) Applied Modelling and Computation Group Earth Science and Engineering Supervisors: Prof C Pain, Prof A Goddard KNOO Post Doc. Support: Dr J Gomes

Contents 1.Brief Description of Generic VHTR 2.Motivation for Modelling 3.Generation of Xsections for Whole Core FETCH Analysis using WIMS9 4.Determination of Reactivity Coefficients 5.RZ Whole Core Transient Example using FETCH 6.Future Aims

What is a VHTR? Evolutionary HTGR for higher coolant temperatures Combined electricity generation process heat applications (Hydrogen) HTGR well established design, handful prototype/demonstration reactors - DRAGON (UK) - AVR, THTR (Germany) - Peach Bottom, Fort St Vrain (USA) - HTTR (Japan) - HTR-10 (China) Current Inter/national V/HTR programs  PBMR,GT-MHR,GTHTR,HTR-PM ANTARES, NGNP Decommissioned Operational

What is a VHTR? Thermal nuclear reactor classified by choice of fuel, moderator coolant Graphite moderated, helium cooled, TRISO fuel with epi-/thermal spectrum Possible for flexible fuel cycle (initial design with U “open” cycle) - THTR (Germany)  Thorium - GT-MHR (Russia)  Plutonium Economics of scale  Economics of repetition (Modular) Strong emphasis on Inherent/passive safety Direct/Indirect Brayton/Rankine/Combined high efficiency (>45%) cycle Modular, Simplicity of Design  Less capital investment High Burn up ~ 150 MWd/Kg Uranium Helium coolant ~ 1000 C

What is a VHTR? Ref. G. Lohnert, “How to obtain an inherently safe HTR”, Raphael HTR Course, 2007 TRISO – Triple Isotropic coated particle All current V/HTR concepts designed around this coated particle concept Primary defence against release of FP Carbon Buffer Layer PyC Layer SiC Layer Fuel Kernel – U, PU, TH Ratio Clad:fuel much higher than LWR

What is a VHTR? - Cylindrical - Annular

VHTR Inherent/Passive Safety Features Negative temperature coefficient  natural shutdown during power excursion Graphite moderated  longer neutronic transient time scales (more collisions) Slow core temperature rise  graphite provides large thermal inertia Helium cooled – chemically and neutronically inert, single phase TRISO particle retaining fission products to high temperatures ~ 1600 C Passive removal of decay heat via natural processes (conduction, convection and radiation) during primary coolant failure  effective due to low power density Simplicity of design (compared to current LWR’s) due to less reliance on redundant safety systems These are characteristics held by certain HTGR’s and desired for V/HTR conditions of higher outlet temperatures

Motivation for Coupled N-TH Modelling To ensure a safe and reliable design Ascertain core (fuel, RPV) temperatures and neutron fluxes during transients Understand complex coupled physics during transients/steady state Each reactor has a class of accidents called Design Basis Accidents. - P-LOFC, D-LOFC - RIA (control rod ejection) - Water/Steam ingress from primary circuit coolers - ATWS Capturing the relevant physics requires the use of Coupled Neutronic Thermal-Hydraulic codes Best Estimate (FETCH) approach rather than Conservative - improved safety analysis and confidence in results

Whole Core VHTR FETCH Modelling Ref. INEEL/EXT James W. et al, D Cylindrical Full 3D 1/6 3D

Multiscale Generating Cross Sections Cross sections  probability of reaction rate Vary with time, space, neutron energy and neutron direction Represent fine scale heterogeneity in homogeneous core model via smeared FA cross sections (cannot resolve billions of TRISO!!) Use an accurate representation of core to give approximate flux density  spatial smearing and energy condensing Start at smallest scale (TRISO), then build up  Fuel Compact  Fuel Assembly (the Lattice Cell) Cross sections generated by reactor physics code WIMS9 (Serco Assurance) Ref. INEEL/EXT James W. et al, 2004

Multiscale Generation of Cross Sections Approximate TRISO ~1000’s WIMS9 Modules: HEAD  PRES  PROC  RES  PROC  PIP  SMEAR Helium Fuel Compact Graphite

WIMS9 Modules: Smear  Cactus  Smear  Cactus  Smear  Condense Smear Multiscale Generation of Cross Sections

Reactivity Temperature Coefficients (WIMS9) Fuel Kernel Moderator (graphite) TRISO Coatings Average Reactivity Coefficients: Fuel ~ pcm/K Mod ~ pcm/K Coating ~ pcm/K Helium ~ 0 pcm/K (small) for fresh UO2 fuel, certain coefficients may become less negative (or positive) with burn up Inherent Safety Characteristic Reactivity = K – 1 K

RZ Whole Core Transient Example using FETCH Power (illustrated by shortest lived delayed neutron precursor) Solid Temperature, C

RZ Whole Core Transient Example using FETCH Power, WMax Solid Temperature, C

Future Aims Coupled Neutronic Thermal Hydraulic analysis of generic VHTR (Block and Pebble) Challenge inherent and passive safety features (design basis accidents) Benchmark neutronic model with Monte Carlo and experimental data Incorporation into FETCH of system code MACE (British Energy) Code-to-code comparison with PANTHER (British Energy) Improved heat transfer correlations (FLUIDITY) Mulitscale thermal sub model  accurate feedback and temperatures Compare Smeared Sub model Ref. gt-mhr.ga.com

Thank you