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Review of the Recent FAST Project Activities Related to Gen-IV Fast Reactors

K. Mikityuk and FAST reactors group:J. Krepel, S. Pelloni, S. Pilarski

PhDs: A. Epiney, A. Chenu, K. SunMSs: R. Adams, A. Pont Ribas, C. Klauser

postDoc: E. Marova

Paul Scherrer Instituthttp://fast.web.psi.ch/

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OverviewFAST project at PSIInternational projectsFAST code system: recent developments and applications

Main components of the FAST code systemModelling of sodium two-phase flow with the TRACE codeNew x-section generation procedure for the PARCS codeESFR equilibrium fuel cycle analysis with the EQL3D/ERANOS procedureDecomposition of reactivity effects in sodium fast reactorsGFR studiesLFR studies

Near-future plansValidation of the FAST code systemMulti-physics uncertainty and sensitivity analysisComparative analysis of the main Generation-IV fast-spectrum cores

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FAST project conceptThe goal of the project is to develop the ability to provide a unique expert analysis

in neutronics, thermal hydraulics and fuel behaviour, with the use of a unique

computational tool, integration into international programs and organization of an efficient team,for three Gen-IV fast-

spectrum systems: sodium-, gas- and lead-cooled reactors.

K. Mikityuk

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FAST international projectsGas-cooled fast reactor

EURATOM: FP6 GCFR (finished in 2008), FP7 GoFastR (2010-2012)CEA & GIF: cooperation in different aspects

Lead-cooled fast reactorEURATOM: FP6 EUROTRANS (2005-2009), FP6 ELSY (2006-2009), FP7 LEADER (2010-2012) – neutronics and safety analysis

Sodium-cooled fast reactorEURATOM: FP7 ESFR (2009-2012) – neutronics and safety analysisCEA: cooperation in neutronicsIAEA TWG FR: PHENIX EOL Test Benchmarks – neutronics and thermal hydraulics

FAST

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Point (0D) reactor kinetics

Fuel thermal mechanics

Re-calculation of reactivity

temperatures in fuel elements

System thermal hydraulics

Static neutronics

Reactivity coefficients,kinetics parameters, power distribution

powerin fuel

coolant pressure, temp. & HXC

Dynamic part:at each time step

Base-irradiation fuel behaviour

Temperature distribution, geometry, stress-strain condition

Core thermal hydraulicscoolant pressure, temp. & HXCtemperatures

powerin fuel

powerin coolant

Core orificing

modified reactivity

powerin coolant

Static part:once before...

Input data for transients

3D reactor kineticsPARCS

Fuel thermal mechanicsFRED

Re-calculation of X-sections

temperatures in fuel elements

System thermal hydraulics: TRACE

Static neutronicsERANOS/EQL3D

X-sections and their derivatives,kinetics parameters

powerin fuel

coolant pressure, temp. & HXC

Dynamic part:at each time step

Base-irradiation fuel behaviour: FRED

Temperature distribution, geometry, stress-strain condition

Core thermal hydraulicsTRACE

coolant pressure, temp. & HXCtemperatures

powerin fuel

powerin coolant

Core orificing

modified reactivity

powerin coolant

Static part

Input data for transients

Structure of the FAST code system FAST

K. Mikityuk, et al. Annals of Nuclear Energy, 32, (2005).

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Modelling of sodium two-phase flow with the TRACE code

Purpose: Development of an advanced computational tool for coupled neutronics thermal-hydraulics transient analysis of sodium systemsImplementation of EOSs and physical model

to simulate sodium boilingBenchmark against SIMMER-IIIBenchmark against experiments: Next step: Use of the FAST code system with

the extended TRACE version to perform a coupled TRACE/PARCS 3D neutronics/thermal-hydraulics transient analysis of the Gen-IV SFR

A. Chenu, PhD

Stable boiling: study friction models

Transient boiling: simulation of a LOF in 2D pin-bundle

study of the interfacial tranfermechanisms

A. Chenu, et al. Nuclear Engineering and Design, 239, 11 (2009)

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Motivation: to improve the x-section generation methodology in the FAST code system to take into account the interrelations between reactivity feedbacks and to prepare for implementation of the uncertainty propagation methodology.

The main idea of the new method is to prepare a set of multi-group micro x-sections for each isotope and each core zone in the form of tables for a given range of temperatures and background cross sections (sigma-zeros) with the use of the ECCO cell code (ERANOS).

During the transient simulation the new procedure within the PARCS code is supposed to calculate for current temperature, sigma-zero and nuclear densities for each core zone a set of macro x-sections interpolating micro x-sections from the pre-generated tables.

New x-section generation model for PARCSS. Pelloni

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The model was implemented in the FAST code systemAs a next step, the new model is tested against the “old” x-section calculation model based

on the basic macro x-sections and their derivatives with respect to state variables (fuel temperature, coolant density, core dimensions, etc)Transients in which interrelations between reactivity feedbacks are weak (e.g. ULOF) were

chosen – good agreement obtained

S. PelloniNew x-section generation model for PARCS

Power in ULOF Reactivity in ULOF

S. Pelloni, et al. Proc. of M&C2009

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Equilibrium closed fuel cycle simulation

Is the core composition the same as in previous cycle?

Fuel cycle calculation

Reload the spent fuel from the core

Make reshuffling in the core

Load the fresh fuel in the core

Fuel cooling down,homogenization and partitioning Final storageFission products

and some MAs

Rest of the fuel

Fuel manufacturing Interim storageFeed fuel(e.g. natural U)

These massesare the same

Spent fuel

Fresh fuel

no

Calculation of safety parametersyes Stop

EQL3D

Long-term fuelcooling down

Short-term fuelcooling down

EQL3D – ERANOS-based calculational procedureMulti-batch equilibrium cycles for fast-spectrum cores in 3D geometryDifferent assumptions about feed fuel (e.g. reactors’ own spent fuel)

J. Krepel

Equilibrium closed fuel cycle analysis

J. Krepel, et al. Annals of Nuclear Energy 36 (2009)

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ESFR equilibrium cycle analysisJ. Krepel

1400

1500

1600

1700

1800

1900

2000

0 410 820 1230 1640 2050Time (EFPD)

Void

reacti

vity (

PCM)

Open cycle: FirstOpen cycle: EQLClosed cycle: FirstClosed cycle: EQL

-1000

-900

-800

-700

-600

-500

0 410 820 1230 1640 2050Time (EFPD)

Dopp

ler co

nstan

t (PCM

)

1500

1600

1700

1800

1900

2000

Void

reacti

vity (

PCM)

Dopp. constant: nominal coreDopp. constant: voided coreVoid reactivity: nominal temp.Void reactivity: Dopp. temp.

One of the SFR preliminary oxide fuel cores designed at CEA

The aim is to simulate and confirm the SFR capability for closed fuel cycle, and to evaluate and compare the parameters of the open and closed cycle at equilibrium

Decomposition of the void reactivityeffect to understand the deterioration of thevoid effect in closed cycleInterrelation between the Doppler and

void effect� new x-section generationmodelSimilar study is under way in frame of

the ESFR project

J. Krepel, et al. Proc. of ICAPP-09 (2009)

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Decomposition of reactivity effects in SFRK. Sun

1.E-061.E-051.E-041.E-031.E-021.E-011.E+00

1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07Energy (eV)

Norma

lized F

lux/Le

thargy

Nominal SpectrumVoided Spectrum

-800

-400

0

400

800

1200

234567891 01 11 21 31 41 51 61 71 81 92 02 12 22 32 42 52 62 72 82 9Energy Group

Reac

tivity

(pcm

)

CaptureFissionProduction

Methodology to better understand the reasons of the positive sodium void effectClosed equilibrium fuel cycle conditionsBoth neutron balance method and perturbation theory algorithmsReaction-wise, isotope-wise, and energy-group-wise decomposition

K. Sun, et al. Proc. of PHYSOR-2010 (2010)

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Gas-cooled Fast Reactor studiesFAST

Contribution to EURATOM projects (FP6 GCFR, FP7 GoFastR) – neutronics and safety analysis

3D analysis of gas-cooled fast reactor core behaviour in control assembly withdrawal accidents (PhD study)

Development and application of an advanced fuel model for the safety analysis of the Gen-IV gas-cooled fast reactor (PhD study)

Heavy gas injection in the Gen-IV gas fast reactor to improve decay heat removal under depressurized conditions (PhD study)

Preliminary design of a Brayton cycle for an autonomous decay heat removal in the Gas Fast Reactor (PhD study)

Calculational investigation in support of the low-temperature gas cooled fast reactor design (postDoc study).

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GFR fuel modelMaterial properties database compiled:

(U-Pu)C for fuel, SiC for matrix2D thermo-mechanical model developed and

validated against 3D FEM analysisTransient analysis performed with new modelLimitations of the new 2D model identifiedDetailed 3D FEM analysis resulted in

recommendationsfor use of 1D models;for fuel design optimization

FAST is unique numerical tool for GFR transient analysis with links between fuel behaviour, neutron kinetics and thermal hydraulics

X-Y-Z

X

R-Z

P. Petkevich, PhD

P. Petkevich, et al. Annals of Nuclear Energy 34 (2007)

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Brayton cycle for DHR in GFRIdea: use of decay heat itself to evacuate it. As long as there is decay heat to evacuate

there is energy to assure the cooling � Enhance passivity of DHR at low pressure

Chosen design point:Equilibrium at 2 bar0MW power productionHe mass flow: 32 kg/sTurbine diameter: 1.6 m

A. Epiney, PhD

Pressure guard containment1 – reactor vessel2 – main HX3 – DHR HX4 – gas reservoirs (helium / heavy gas)

A. Epiney, et al. Proc. of NURETH (2009)

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Lead-cooled Fast Reactor studiesK. Mikityuk

Contribution to EURATOM projects (FP6 ELSY, FP7 LEADER) – neutronic and safety analysis

Analysis of heat transfer to liquid metal

Modelling of the oxide layer build-up on the surfaces of the structural material in the lead flow with the TRACE code

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Analysis of Heat Transfer to Liquid Metal

100 1000Peclet number

0

5

10

15

20

25

30

35

40

45

Nuss

elt n

umbe

r

Maresca, x=1.75Gräber, x=1.25Gräber, x=1.6Gräber, x=1.95Borishanski, x=1.1Borishanski, x=1.2Borishanski, x=1.3Borishanski, x=1.4Borishanski, x=1.5Zhukov, x=1.25Zhukov, x=1.28Zhukov, x=1.34Zhukov, x=1.46

a + b Pec

0 10 20 30 40 50Measurement

0

10

20

30

40

50

Calculation

Maresca, x=1.75Gräber, x=1.25Gräber, x=1.6Gräber, x=1.95Borishanski, x=1.1Borishanski, x=1.2Borishanski, x=1.3Borishanski, x=1.4Borishanski, x=1.5Zhukov, x=1.25Zhukov, x=1.28Zhukov, x=1.34Zhukov, x=1.46

658 data points for different liquid metalstube bundles with x = 1.1 — 1.95Peclet 30 to 5000

8 correlations analysednew correlation derivedNu = 0.047(1-e-3.8(x-1))(Pe0.77+250)

Recommendations for EU projects ELSY, EUROTRANS and ESFR producedK. Mikityuk, Annals of Nuclear Energy (2009)

K. Mikityuk

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Near-future plansFAST

To focus on sodium-cooled fast reactor with preservation of activities on GFR and LFR for comparison purposes

Three main directions:Validation of the FAST code system

Static neutronics: validation of ERANOS,,e.g. against PHENIX-EOL/control rod withdrawal test data and against BFS/BN-600 data.

Thermal-hydraulics: validation of TRACE, e.g. against PHENIX-EOL/natural circulation test data and against sodium boiling data (KNK/FzK, ISPRA, CABRI/SCARABEE).

Fuel behaviour: validation of FRED, e.g. against HALDEN UO2 base-irradiation data and against CABRI/SCARABEE fast transient data.

Coupled transient simulation: validation of ERANOS/PARCS/TRACE/FRED, e.g. against first phase of the PHENIX-EOL/natural circulation test data and against EBR-II LOF WS and LOHS WS tests.

Multi-physics uncertainty and sensitivity analysisComparative analysis of the main Generation-IV fast-spectrum cores

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Uncertainty propagation methodologyTo develop uncertainty

propagation methodology in frame of the FAST code systemfor comparative evaluation of

the uncertainties of different natures (in neutron data, material properties, models, etc.)for steady-state and transient

analysis of the Gen-IV systems (SFR first)

FAST

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Three Gen-IV systems – multibatch schemesJ. Krepel

SFR3600MWth

2-zones core

5-batch subdivision

GFR2400MWth

2-zones core

3-batch subdivision

LFR1500MWth

3-zones core

3-batch subdivision

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Comparison of equilibrium safety parameters

-1300-1200-1100-1000

-900-800-700-600-500

0 277 554 831 1108 1385 1662 1939 2216 2493Load (EFPD)

Dopp

ler co

nstan

t

SFR: voided conditionsSFR: nominal conditionsGFR: voided conditionsGFR: nominal conditionsLFR: voided conditionsLFR: nominal conditions

Doppler constant, pcm Beta effective, pcm

340342344346348350352354356358360

0 277 554 831 1108 1385 1662 1939 2216 2493Load (EFPD)

Beta

effe

ctiv

e

LFR: nominal conditionsSFR: nominal conditionsGFR: nominal conditions

SFR – strong dependence of Doppler on void conditionsGFR – the best DopplerBeta effective very close for all systems

J. Krepel

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Comparison of equilibrium safety parameters

Void reactivity is calculated assuming the voidage of the fuel region only

Core: SFR GFR LFR Strategy open

cycle closed cycle *

open cycle

closed cycle *

open cycle

closed cycle *

Residence time (EFPD)

5 X 410 =2050

5 X 410 =2050

3 X 831 =2493

3 X 831 =2493

3 X 831 =2493

3 X 831 =2493

Reactivity swing 773 642 3139 1265 1134 580 Pu content (w%) 17.7 17.9 18.1 18.8 16.5 16.9

MA content (w%) 0.5 1.0 0.3 1.3 0.3 1.2

Doppler constant -898 -15% -1326 -9% -851 -15%

Void reactivity 1775 +6% 215 +23% 253 -6%

β 370 -6% 368 -6% 370 -6%

Λ (ms) 0.41 -10% 0.98 -10% 0.86 -15%

* closed cycle = 100% Actinides recycling by reprocessing & using of natural uranium feed

J. Krepel, et al. Proc. of ICAPP-10 (2010)

J. Krepel

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ConclusionsThe FAST code system is a computational tool that gives us unique

opportunity to compare the breeding and safety parameters for the main Gen-IV fast-spectrum system in equilibrium open and closed cycleThe one of the important current R&D directions is the coupled 3D

neutron kinetics/thermal-hydraulic analysis of the SFR transients with sodium boilingThe important future directions include further validation of the code

system, continuation of the SFR/GFR/LFR comparison study and implementation of the multi-physics uncertainty/sensitivity methodology for static and transient analysis

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Thank you! Any questions?

FAST