Coupled Neutronic Fluid Dynamic Modelling of a Very High Temperature Reactor using FETCH

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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

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Contents

1. Brief Description of Generic VHTR
2. Motivation for Modelling
3. FETCH Model
4. FEM based Dual Submodel
5. Step Reactivity Transient Results

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VHTR Description

• 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
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VHTR Description
TRISO Triple Isotropic coated particle
All current V/HTR concepts designed around this
coated particle concept Primary defence against
release of FP
PyC Layer
SiC Layer
Carbon Buffer Layer
Ratio Cladfuel much higher than LWR
Ref. G. Lohnert, How to obtain an inherently
safe HTR, Raphael HTR Course, 2007
Fuel Kernel U, PU, TH
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VHTR Inherent/Passive Safety Features
TRISO fueled (high temperature burnup capable)
Negative temperature reactivity coefficient for
fuel
Low Core Power Density 6MW/m3
Large thermal inertia of graphite core
Helium Cooled (single phase, chemically and
neutronically inert)?
Passive decay heat removal (conduction,
convection, radiation)?
Simplicity of design
Ref. INEEL/EXT-03-00870 NGNP, 2003
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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
• - asymmetric 3D transients with
  local effects (fine detail)?

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FETCH Model
TRISO coated fuel particle 1mm diameter
80cm
36cm
Block Type Fuel Assembly
Ref. INEEL/EXT-04-02331 NGNP, 2004
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FETCH Model
Full 3D
Ref. INEEL/EXT-04-02331 NGNPl, 2004

• Homogenised - Fuel Assemblies
• - Control Rod
• - Burnable Poison
• Fluids --gt Multi-phase porous media
• Radiation --gt Smeared group constants

2D Cylindrical
1/6 3D
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FETCH Model
WIMS9
Tabulated Group Constants
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FETCH Model
EVENT
FLUIDITY
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Dual Submodel

• 1D FEM time dependent submodel implemented into
  FETCH.
• Within each FETCH fuel element a submodel is
  solved (heat conduction equation) based on the
  characteristic heterogeneous geometry
• Currently fuel compact with gap only in submodel
  and global solid phase is surrounding moderator
• Linked through BC of submodel
• Same method used for subsubmodel within submodel
  (TRISO within fuel compact)?
• Neutronic group constants generated for 2D grid
  of varying fuel and moderator temperature
• includes explicit cross terms

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Generating Group Constants (WIMS9)?
Group Constant
TRISO Coatings
Fuel Kernel
Sg
Reactivity
Reactivity
Inherent Safety Characteristic
Reactivity Temperature Coefficient
Moderator (graphite)?

• for fresh UO2 fuel, certain coefficients may
  become less negative (or positive) with burn up

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Whole Core RZ

• Annular fuel region with above/below reflector
  solved for coupled fluids neutronics
• Inner/outer reflector solved for just neutronics
• Helium coolant flows down from top of annular
  core domain with inlet temperature of 500 degrees
• Burnup control inserted, startup control absent,
  start zero power
• Solid annular fuel region has start temperature
  of 800 degrees
• Reactivity insertion transient initiated through
  core cooling
• No automated intervention (SCRAM,TRIP) during
  transient
• Account double heterogeneity for heat transport
  by double sub-model using time dependent FEM
• -- subsub-model within sub-model within global
  FETCH element
• Fuel cross sections generated as function of
  fuel and moderator temperature including cross
  terms using WIMS9

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Whole Core RZ

• Core initially sub critical
• Core criticality induced by coolant removing
  energy from system

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Whole Core RZ

• Fuel core starting temperature now 700 C
• Much more energetic transient

• Core started in delayed critical state

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Whole Core RZ
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Whole Core RZ
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Whole Core 3D

• Step Reactivity insertion delayed critical
• Constant inlet BC's --gt run to steady state
• 50,000 elements
• P1 angular approximation

Thermal Flux
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Whole Core 3D
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Whole Core 3D
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Acknowledgements

• AMEC NNC for providing helpful advice

Contact

• brendan.tollit05_at_imperial.ac.uk
• c.pain_at_imperial.ac.uk
• amcg.ese.ic.ac.uk

Thank you! Questions?