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