coupled neutronic fluid dynamic modelling of a very high temperature reactor using fetch brendan...
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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
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-04-02331 James W. et al, 2004
2D 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-04-02331 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 ~ -6.965 pcm/K
Mod ~ -0.704 pcm/K
Coating ~ -0.0475 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, W Max 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