performance evaluation of metallic fuel for sfr · 1 fr09, kyoto, 7-11 december 2009 performance...

1 FR09, Kyoto, 7-11 December 2009

Performance Evaluation of Metallic Fuel for SFR

International Conference on Fast Reactors and Related Fuel Cycles (FR09), Kyoto, Japan

9 December 2009

Byoung Oon LEE


Outline

IntroductionIDiffusion Couple Test IIIrradiation TestIIIModelingIVMetal Fuel DesignVSummaryVI


IntroductionI


Introduction� Metal fuel development project for SFR in Korea

– U-TRU-Zr metallic fuel– Cladding material : FMS� The performance analysis is essential to assure an

adequate fuel performance and its integrity� The present study represents progress results of evaluating

the performance of metal fuel for SFR in Korea ’07 ’10 ’15 ’20 ’25 ’28

Metal Fuel AssemblyFab. (U-X-Zr)

Fuel Fab. In Hot Cell(U-TRU-Zr)

Fuel Fab. Facility Operation(U-TRU-Zr)

Fuel Fab. for Demo. SFR(U-TRU-Zr)

Metal Fuel RodFab. (U-X-Zr)Fuel

Develop.

Fuel Perform.

Eval.Fuel Rod Irradiation in Fast Reactor (Abroad)

Fuel irradiation in HANARO

Verification Irradiation of Fuel Assembly (Abroad)Transient Tests in Hot Cell Transient Behavior Tests in Reactor (Abroad)

Fuel Design, Fuel Performance Modeling, and Computer Code Development

Control Rod Assembly Development


Diffusion CoupleDiffusion CoupleDiffusion CoupleDiffusion Couple

TestsTestsTestsTestsIrradiation TestIrradiation TestIrradiation TestIrradiation Test

ModelingModelingModelingModeling

Fuel Performance Evaluation

�Verification of Fuel PerformanceVerification of Fuel PerformanceVerification of Fuel PerformanceVerification of Fuel Performance

�Design and fabrication of Design and fabrication of Design and fabrication of Design and fabrication of

Irradiation Capsule in HANAROIrradiation Capsule in HANAROIrradiation Capsule in HANAROIrradiation Capsule in HANARO

�Diffusion couple test between Diffusion couple test between Diffusion couple test between Diffusion couple test between

metal fuel and cladding metal fuel and cladding metal fuel and cladding metal fuel and cladding

� UUUU----ZrZrZrZr----X vs. HT9 or T91 or X vs. HT9 or T91 or X vs. HT9 or T91 or X vs. HT9 or T91 or

Barrier claddingBarrier claddingBarrier claddingBarrier cladding

�Thermal Conductivity Thermal Conductivity Thermal Conductivity Thermal Conductivity

�FGRFGRFGRFGR

�Fuel constituent migrationFuel constituent migrationFuel constituent migrationFuel constituent migration

�Deformation etcDeformation etcDeformation etcDeformation etc

Metal Metal Metal Metal Fuel Fuel Fuel Fuel

Performance Performance Performance Performance

EvaluationEvaluationEvaluationEvaluation

10

15

20

25

30

35

40

0 100 200 300 400 500 600 700 800Temperature, °C

Therm

al co

nduc

tivity

, W/m

-K

U-10Zr(measured)U-Zr-Ce(measured)U-10Zr[Bill one]U-10Zr[Takahashi]U-Zr-Ce(evaluated)

10

15

20

25

30

35

40

0 100 200 300 400 500 600 700 800Temperature, °C

Therm

al co

nduc

tivity

, W/m

-K

U-10Zr(measured)U-Zr-Ce(measured)U-10Zr[Bill one]U-10Zr[Takahashi]U-Zr-Ce(evaluated)

T91 V U-10Zr-4CeT91 V U-10Zr-4Ce

FMFMFMFM

SSSS

Cr

Cr

Cr

Cr22 22OO OO

33 33

UUUU----

10Zr10Zr10Zr10Zr

FMFMFMFM

SSSS

Cr

Cr

Cr

Cr22 22OO OO

33 33

UUUU----

10Zr10Zr10Zr10Zr

Fuel Design Fuel Design Fuel Design Fuel Design

�Evaluation of Design Limit by Evaluation of Design Limit by Evaluation of Design Limit by Evaluation of Design Limit by

MACSIS etcMACSIS etcMACSIS etcMACSIS etc

�Fuel Spec for High Fuel Spec for High Fuel Spec for High Fuel Spec for High burnupburnupburnupburnup and and and and

high temperature operation high temperature operation high temperature operation high temperature operation

0.0000

0.0005

0.0010

0.0015

0 2 4 6 8 10 12 14

Burnup (at%)

CDF

Inner fuel, Mod. HT9,1.75 plenum

middle fuel, Mod. HT9,1.75 plenum

outer fuel, Mod. HT9,1.75 plenum

middle fuel, Mod. HT9, 2.0 plenum

outer fuel, Mod. HT9,2.0 plenum

CDF limit


Metal Fuel Configuration


Diffusion Couple Test II

I.1 Diffusion couple tests without barrierI.2 Diffusion couple tests with a metallic foil

barrierI.3 Diffusion couple test with a surface

treatment


II.1 Diffusion couple tests without barrier�Diffusion couple tests of U-Zr-(0, 2)Ce with FMS (700℃℃℃℃)

– Various interaction phases– Major phase : UFe2, U6Fe, and ZrFe2– Similar results of US and JAPAN

� Diffusion couple tests of U-Zr-(0, 2)Ce with FMS (740℃℃℃℃)

– Thickness of the interaction region : different from Keiser’s observation

• microstructures were similar – Gray phase of UFe2, dark one of Zrrich-line layer, and mixed phase with U and Zr were observed

U–10Zr vs. HT9 (740℃℃℃℃, 96 h)

U–10Zr vs. HT9 (700 ℃℃℃℃, 96 h)

HT9 U-10Zr

Zr richUFe2

(U,Zr) phase


II.1 Diffusion couple tests without barrier� U-10Zr and HT9 (800℃, 25h)

–SEM revealed that the clad lost about 250㎛ of its thickness–From the EDX

• U, Fe and Cr diffused into each other at the opposite direction �U-Zr-Fe-Cr compound as U6(Fe,Cr), Zr(Fe,Cr)2, and U(Fe,Cr)2

Fuel CladInterface

Diffusion of Fe, Cr

Diffusion of U, Zr

Fuel CladInterface

Diffusion of Fe, Cr

Diffusion of U, Zr

U–10Zr vs. HT9 (800℃℃℃℃, 25 h)


II.2 Diffusion couple tests with a metallic foil barrier

� Diffusion couple tests of U-Zr-(0, 2)Ce were carried out for the barrier foils –Zr, Mo, Nb, Ti, Ta, V, and Cr

� Zr–dissolved into the matrix – its thickness was significantly reduced � Mo

–part of the Mo element reacted with the U-10Zr fuel� Nb, Ti

–barrier elements and the fuel diffused into each other so that the reaction layer was formed.

� Ta–EDX analysis revealed that Ta reacted with the fuel, where it diffused into the fuel component.


II.2 Diffusion couple tests with a metallic foil barrier� V

– no reaction between the barrier material and the fuel component• Inter-diffusion was completely prevented

– EDX analysis revealed that there was no U in the V layer.

– V–Fe–Zr layer was observed between the V foils and the FMS • measured composition : 93.5 at.%V–4 at.%Fe–2.5 at.%Zr.

• Reduction of V foil thickness� Cr

– Neither inter-diffusion nor eutectic reaction

� V and Cr exhibited the most promising performance

U–10Zr and HT9 with a V barrier foil (800℃℃℃℃, 25 h)

Zr

V

93.5V-4Fe-2.5Zr

HT9 U-10ZrV

Summary of the metallic foil barrier performance

NoNoYesCrNoNoYesVYesNoYesTaYesNoYesMoYesYesYesTiYesYesYesNbYesYesYesZr

Reaction with fuel

Element Interdiffusion

Eutectic melting preventionElement


II.3 Diffusion couple test with a surface treatment�Final application of a barrier

cladding requires the barrier in the surface of a cladding tube –Electroplating (for Cr) and vapor deposition (for Zr and V)

�Cr barrier electroplated on FMS–Cr prevented the eutectic melting between U-10Zr and HT9 at 700, 740, and 800 ℃℃℃℃. –Fuel constituent such as U have penetrated locally along the crack in the barrier

• uranium compound along the clad surface

U-10Zr and HT9 with the electroplating of Cr after the

diffusion couple test (800℃℃℃℃, 25 h)

� It seems that a crack affects the fuel-cladding interaction in a negative way– further analysis will be carried out to investigate the exact phenomenon


II.3 Diffusion couple test with a surface treatment�Zr-vapor-deposited barrier

–no visible phase formation –excellent barrier performance, contrary to the case of the Zrmetallic foil– further analysis will be continuously carried out

�V-vapor-deposited barrier–not effectively prevent inter-diffusion contrary to the V foil– It seems that the thickness of V (~1.3 μμμμm) was too thin –Diffusion couple test will be carried out with thick barrier cladding

U-10Zr vs. HT9 with the Zr vapor deposition (800℃℃℃℃, 25 h)

ZrZr

ZrCr

Fe

V

Zr Cr Fe

Fe

Cr

U-10Zr vs. HT9 with the V vapor deposition (800℃℃℃℃, 25 h)


Irradiation TestIII


Fuel irradiation test � Schedule

– 2010 : irradiation in HANARO– 2008-2009 : irradiation capsule design and fabrication

� Objectives – Identify the Ce-bearing fuel performance and the characteristics of barrier cladding

� Irradiation condition– Fuel : U-10Zr and U-10Zr-6Ce– Cladding : FMS– Maximum burnup : 3 at% (1st HANARO irradiation test) – Linear power : 306 W/cm, – Expected duration : ~150 EFPD

� Fast reactor condition – Thermal neutron flux filter : Hf or Boral plate– Fuel temp. control : He gap– Fuel/cladding gap bonding :Na

Fuel

Na

Outer tube

Gap (He, 60µm)

Coolant

Cladding

Fuel Slug Cross Section

Upper Part

Na

Fuel Slug Cross Section

Upper Part

Na


Irradiation capsule � Capsule design

– Two test sections• each section accommodates six

rodlets– The fuel rod is contained in the

sealing tube for safety in case of sodium leakage from the cladding

Mid-core

bottom top

Mid-core

bottom top

flux

Lower partUpper part

Irradiation capsule dimension

5.628.624.65.515.83.7

Fuel Cladding Outer tube Dia. (mm) Density (g/cm3) Outer dia.(mm) Inner dia.(mm) Outer dia.(mm) Inner dia.(mm)


ModelingIV

IV.1 Thermal conductivity modelIV.2 Fuel constituent migration model


MACSIS CODE

INPUT Geometry, Power History, Model Specifications

Calculate Initial Conditions

Power, Burnup, Fluence

Coolant Temperature

Cladding Temperature

Cladding Deformation

Fuel/Clad HTC

Fuel Temperature

Constituent Redistribution

Fuel Swelling

Fission Gas Release

Plenum Pressure

Failure Probability

OUTPUT

Is time ended?

No

F/C Gap Converge?

No

Yes

Yes

STOP

� Main structure� Fuel temp. calculation routine� FGR calculation routine� Cladding deformation calculation routine

� Main Function� Axial and radial temperature distribution� Fission gas release including He release� Fuel constituent migration� Cladding deformation by plenum pressure� Cumulative damage fraction (CDF) � Probabilistic estimation of CDF

in connection with Weibull analysis � Cladding wastage effect by eutectic melting� FCMI by solid fission product

� MA (+ RE) bearing metal fuel behaviormodel is now developing

MACSIS Flow Chart


IV.1 Thermal conductivity model�RE is precipitated in the U-Zrmatrix

– cause a variation of fuel properties �The effect of the Ce precipitates on effective thermal conductivity is evaluated

– based on models for heterogeneous materials such as Maxwell and Bruggeman models

�Thermal conductivity of the U-Zralloy is reduced

– with an increasing Ce content�Addition of Ce up to 6 wt%

– reduce the thermal conductivity of the U-Zr alloy by less than 5%.

• low volume fraction of the Ce phase• relatively high thermal conductivity of Ce

Effect of the Ce content on the thermal conductivity of U-10Zr

0.95

0.96

0.97

0.98

0.99

1.00

0 200 400 600 800Temperature, °C

kU-10Z

r-Ce /k

U-10

Zr

2Ce

4Ce

6Ce


IV.1 Measured thermal conductivity �The measured thermal conductivity of the U-Zr-Ce alloy

– by the product of the thermal diffusivity, density, and specific heat

• Thermal diffusivity : laser flash method

• Density : immersion technique• specific heat : Kopp-Neumann’s law

�Thermal conductivity of U-Zr lies in between the Billone or Takahashi’s evaluation

– The effect of Ce on the thermal conductivity of U-Zr alloy is well described by the present heterogeneous mixture models

Comparison of the thermal conductivities for U-Zr-Ce alloys

10

15

20

25

30

35

40

0 100 200 300 400 500 600 700 800Temperature, °C

Therm

al co

nduc

tivity

, W/m

-K

U-10Zr(measured)U-Zr-Ce(measured)U-10Zr[Billone]U-10Zr[Takahashi]U-Zr-Ce(evaluated)


IV.2 Fuel constituent migration model�U-Pu-Zr Migration

– Based on the Ishida’s model and Hofman’s theory – Reconstruct the quasi-binary U-Zrphase diagram by Ishida’s Concept– Assumption of the diffusion coefficient by Hofman’s theory�Am migration model for U-TRU-Zr

– by using the U-Pu-Zr migration model �The radial profile of Zrredistribution

– The main reason for the migration is the radial solubility change of Zr– The heat of transport also plays an important role in the migration– depletion of Zr in the middle zone was simulated – value of the heat of transport : more than -100,000kJ/mole.

Radial distribution profile of Zr

0

0.1

0.2

0.3

0.4

0.5

0.6

0 0.2 0.4 0.6 0.8 1

Slug Radis (R/R0)

Zr A

tom

Fra

ctio

n

EBR-II data by Ishida EBR-II data by Ishida EBR-II data by Ishida EBR-II data by Ishida initial Zr concentrationinitial Zr concentrationinitial Zr concentrationinitial Zr concentration

calculated value (Q=-100000)calculated value (Q=-100000)calculated value (Q=-100000)calculated value (Q=-100000)507-694 range507-694 range507-694 range507-694 range542-671 range542-671 range542-671 range542-671 range

initial Am concentrationinitial Am concentrationinitial Am concentrationinitial Am concentration


IV.2 Am migration �Am migration

– calculated by using the MACSIS code and the X501 data

�The migration behavior of Am is similar to that of Zr

– -100,000kJ/mole of the heat of transport was also used

�Simulated Am migration for the X501 fuel

– Am migration along with the migration of Zr was simulated

– At around 700℃℃℃℃ of the fuel centerline temperature, the model predicted that the Am fraction in the fuel center reaches its peak

– There were no centerline Am depletions expected in all range of temperature

Radial distribution profile of Am

0

0.1

0.2

0 0.2 0.4 0.6 0.8 1

Slug Radis (R/R0)

Am

Ato

m F

ract

ion

initial Am concentrationinitial Am concentrationinitial Am concentrationinitial Am concentration

518-722 range (AM)518-722 range (AM)518-722 range (AM)518-722 range (AM)




Metal Fuel DesignV


Design Criteria � No Fuel Melting

–Fuel Temperature ≤ solidus temperature� No Eutectic Melting

–No eutectic liquifaction• Fuel Surface Temperature (TRU% < 19wt%) ≤ 700 ℃• Fuel Surface Temperature (TRU% ≥19wt%) ≤ 650 ℃

� Cladding Limit–Strain limit criteria

• Thermal creep strain : 1%• Total strain : 3%• Swelling : 5%

–Cumulative Damage Fraction (CDF) limit criteria • Steady state operation : 0.001• Transient operation : 0.2


Key design parameters

868940Fuel Slug Length (mm)2.251.75Plenum-to-fuel ratio

1.00/0.72/0.59Cladding Thickness (mm)- Inner/middle/outer

7.09.0Pin Outer Diameter (mm)Mod.HT9Mod.HT9Cladding Material

7575Smeared Density (%)

U-30TRU-ZrU-12.6Pu-0.5Am-0.09Cm-0.06Np-10ZrFuel Slug Contents (wt%)

Preliminary case 1 of SFR conceptual design

KALIMER-600 (case 1 core)


CDF Limit Analysis for KALIMER-600�HT9

– CDF > 0.001 – when the cladding temperature becomes higher than 625 ℃℃℃℃

�Mod.HT9– CDF > 0.001 – when the cladding temperature becomes higher than 645 ℃℃℃℃.

�625 and 645 ℃℃℃℃ are selected as the peak clad temperature

– for the HT9 clad and the Mod.HT9 clad for the Case-1 core, respectively.

CDF as a function of operating temperature for KALIMER-600

(Case-1 core)

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

580 600 620 640 660 680

Temperature(oC)

CDF

HT9

Mod.HT9

CDF design limit


CDF LIMITS FOR SFR Preliminary design�CDF <0.001

– Cladding temperature :650℃℃℃℃– P/F ratio : 2– R/th ratio : 5.5

� If the P/F ration was enlarged to 2.25

– R/th ratio needed to satisfy the CDF limit : 6

� If the plenum-to-fuel ratio was enlarged– It was expected that the Mod.HT9 cladding satisfied the CDF limit at the discharge burnup goal

�Sensitivity analysis according to the design parameter will be performed continuously

CDF according to the cladding temperature, P/F ratio, and R/th ratio

0.0000

0.0010

0.0020

0.0030

0.0040

0.0050

0 2 4 6 8Radius/thickness ratio

CDF

temp 650C, P/F=2temp 650C, P/F=2temp 650C, P/F=2temp 650C, P/F=2temp 650C, P/F=2.25temp 650C, P/F=2.25temp 650C, P/F=2.25temp 650C, P/F=2.25temp 610C, P/F=2temp 610C, P/F=2temp 610C, P/F=2temp 610C, P/F=2


SummaryVI


Summary� Sodium-cooled fast reactor(SFR) is being developed in

combination with the pyro-processing of spent fuel–U-TRU-Zr metallic fuel is a reference fuel for SFR � Fuel performance evaluation is being performed in the

following tasks; – fuel-cladding diffusion couple tests– fuel irradiation test– performance analysis model development– fuel design � This work forms the basis for establishing key technology

that will evaluate the performance of U-TRU-Zr metallic fuel.

performance evaluation of metallic fuel for sfr · 1 fr09, kyoto, 7-11 december 2009 performance...

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