korea atomic energy research institute 2004. 11.09-11 the preliminary performance analysis of the...

Korea Atomic Energy Research Institute

2004. 11.09-11

THE PRELIMINARY PERFORMANCE ANALYSIS OF

THE TRANSMUTATION FUEL FOR HYPER

THE PRELIMINARY PERFORMANCE ANALYSIS OF

THE TRANSMUTATION FUEL FOR HYPER

B.O.Lee, W.S.Park, Y.Kim, T.Y.Song

OECD/NEA 8th Information Exchange Meetings on Actinide and Fission Product Partitioning and TransmutationLas Vegas, Nevada, USA 9 – 11 November 2004


Contents

- Contents-

- Introduction

- Code description

- Design Parameter

- Fuel Temperature Prediction

- Fission Gas Release and He Release Rate Insertion

- Strain Limits Analysis

- Cumulative Damage Fraction Analysis

- Conclusions

2


Introduction

3

U-TRU-(40-60)Zr metallic fuel for HYPER in Korea

HYPER (HYbrid Power Extraction Reactor ) : ADS & sub-critical system

U-TRU-15Zr metallic fuel for critical system

Steady-state performance computer code development

MACSIS-H

(A Metallic fuel performance Analysis Code for Simulating In-reactor behavior under Steady-state conditions)

Parametric study for the selection of nominal design characteristics and operating limits

Material data of U-Pu-Zr : used for those of U-TRU-Zr fuel

Fuel temperature distribution : check the operational limits

Constituent migration analysis : by the quasi-binary U-Zr model

He production rates : inserted into the swelling/FGR routine

Burnup limits : derive the design concept

CDF(cumulative damage fraction) : estimate the failure probabilities


MACSIS-H Code Description

4

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 - Swelling and FGR calculation routine - Cladding deformation calculation routine

● Main Function - Axial and radial temperature distribution - Fuel slug swelling - Fission gas release including He release - Fuel constituent migration - Cladding deformation by plenum pressure - Cumulative damage fraction by the input of other

program - Cladding wastage effect by eutectic melting (now

developing) - FCMI by solid fission product (now developing)

MACSIS-H Flow Chart


Design Parameter

5

Sub-critical (HYPER) Critical

Fuel Slug Contents (wt%) 8U-34.3Pu-4.2Am-1.6Cm-1.7Np-0.2RE-50Zr

65.5U-18Pu-0.5Am-0.4Cm-0.5Np-0.5RE-14.6Zr

241Am Content (wt%) 2.66 0.1

Fuel Slug Diameter (mm) 5.63 6.63

Smeared Density (%) 75 75

Pin Outer Diameter (mm) 7.7 8.8

Cladding Thickness (mm) 0.6 0.57

Fuel Slug Length (mm) 1,500 1,240

Peak Linear Power (kW/m) 28.5 28.9

Coolant Outlet Temperature (°C) 490 540

Cladding Material HT9 HT9

● Key Design Parameter


6

● Transmutation Fuel Composition (w/o) in an Equilibrium Cycle

HYPER

Isotope Feed Charge Discharge

U-234U-235U-236U-238

Np-237Pu-238Pu-239Pu-240Pu-241Pu-242Am-241

Am-242mAm-243Cm-242Cm-243Cm-244Cm-245Cm-246

ReFP*

1.7E-30.0600.0406.466.252.09

46.9620.804.004.246.73

0.0161.12

3.8E-50.0030.22

0.0230.0031

0.980.00

0.680.180.32

15.353.354.89

23.1328.104.228.165.310.302.71

0.0180.0191.910.610.380.370.00

0.620.170.31

14.152.174.36

14.3024.083.757.363.760.302.510.21

0.0181.990.610.383.72

15.25

Critical

Isotope Feed Charge Discharge

U-234U-235U-236U-238

Np-237Pu-238Pu-239Pu-240Pu-241Pu-242Am-241

Am-242mAm-243Cm-242Cm-243Cm-244Cm-245Cm-246

ReFP*

0.0170.600.40

64.802.090.70

15.706.951.341.422.25

0.00530.37

1.28E-50.00100.075

0.00780.0011

3.280.00

0.170.150.43

75.240.620.87

10.487.360.831.440.10

0.0560.43

0.00200.0020

0.260.0790.0450.550.00

0.150.0870.39

68.860.410.798.926.650.761.300.71

0.0560.39

0.0270.0021

0.270.0790.0452.267.84

* Without RE

Design Parameter


7

500 550 600 650 700 750500

550

600

650

700

750

MACSIS Prediction Linear Fitting Line

MA

CS

IS P

redic

tions (

o C)

LIFE-M, etc. Predictions (oC)

Fuel Temperature Prediction

● Fuel Temperature Prediction

Thermal conductivity for the unirradiated U-TRU-Zr alloy by Billone et al.

2620 TWp7.211038.9TWp9.0

Wz61.11

Wz061.011054.1Wp62.2

Wz61.11

Wz23.215.17k

Thermal conductivity for more than 20wt% of Zr or more than 30wt% TRU by AAA Report (LA-UR-02-2630)

CABA

CABACABACACABABACBA kk

kkvvvkvkk

422

The porosities and sodium infiltration effect by Bauer et al

PP

P

k

k

k

k

PfNa

g

Na

Na

g

1 31

1

23

210

0

3 2

( )( ) /

The predicted temperatures of U-Zr metal fuel by MACSIS-H versus LIFE-M by Billone, Pahl, Hofman et al.

reasonably good capability in predicting fuel pin rod temperatures


● Fuel temperature limits on fuel melting

calculated solidus temperature of U-42TRU-50Zr and U-20TRU-14.6Zr : 1090 and 1295oC, respectively

calculated power-to-melt for HYPER fuel: 420 W/cm

(in case of 50 % degraded thermal conductivity and at hot channel)

calculated power-to-melt for critical system fuel : 500 W/cm

8

The operating limits on linear power rate for metallic fuel pin in HYPER and critical system

400

600

800

1000

1200

1400

200 300 400 500 600 700

Linear Power, W/Cm

Tem

p. (

o C)

fuel center at k=0.5ko(HYPER)fuel center at hot channel and k=0.5ko(HYPER)fuel center at hot channel and k=0.5ko(critical)fuel center at k=0.5ko(critical)

power-to-meltfor HYPER

power-to-meltfor critical

melting temp. for U-42TRU-50Zr

melting temp. for U-20TRU-14.6Zr



9

J D CQ C

RTT

1 1 1

1 1

2

~*

TRT

CQ

r

CCDJ S

21

*111

~

11 2/

J D C VH Q

RTT

S f

S

1 1 1

1 1

2

~ *

● Fuel Constituent Migration

Based on the Ishida’s model and Hofman’s theor

y

Reconstruct the quasi-binary U-Zr phase diagra

m by Ishida’s Concept

Assumption of the diffusion coefficient by Hofm

an’s theory

Flow Chart of Calculation Scheme

Diffusion equation for constituent migration

Single phase

Multi-phase

Boundary



● Calculated and measured radial profile of Zr for the U-19Pu-10Zr

driving forces acting on the Zr migration : molar enthalpy of solution ( Hs△ ), heat of transport (Q*) , and c

oncentration gradient

The main reason for redistribution : radial solubility change of Zr

The heat of transport plays a role in the redistribution

discrepancy was small : in case of Q* : -97,000kJ/mole

significant amount of Zr is depleted in the middle zone

0

0.1

0.2

0.3

0.4

0 0.2 0.4 0.6 0.8 1

Slug Radis (R/R0)

Zr A

tom

Fra

ctio

n

EBR II experimental datainitial Zr concentrationcalculated value (Q=-50000)calculated value (Q=-77000)calculated value (Q=-97000)

680ºC 2 + + 510ºC

0

0.1

0.2

0.3

0.4

0 0.2 0.4 0.6 0.8 1

Slug Radis (R/R0)

Zr A

tom

Fra

ctio

n

EBR II experimental datainitial Zr concentrationcalculated value (Q=-50000)calculated value (Q=-77000)calculated value (Q=-97000)

680ºC 2 + + 510ºC

10

Calculated and measured radial profile of Zr for U-19Pu-10Zr



11

● Calculated radial profile of Zr for the U-20TRU-14.6Zr

Zr fraction in fuel center at around 670 oC : 0.5

fuel centerline melting : retarded by addition of Zr

Sharp Zr depletion at the upper limit of (+) phase boundary

melting temperature and eutectic-melting point decreased

This phenomenon has not been confirmed yet experimentally

Calculated Radial profile of Zr for U-20TRU-15Zr

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Slug Radius(R/Ro)

Zr A

tom

ic F

ract

ion

Zr redistribution at Z=2/3ZoZr redistribution at Z=1/3Zoinitial Zr concentration

618oC

748oC

669oC

528oC



● Margin to slug centerline melting

Centerline temperature : very lower than the solidus temperature

Sufficient Margin to the slug centerline melting

→ same for the case of the fuel constituent migration

12

radial temperature for U-20TRU-14.6Zr

500

600

700

800

900

1000

1100

1200

1300

1400

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Distance from the Fuel Center (r/Ro)

Fuel

Slug

Tem

pera

ture

(o C)

Constituent Migration + Sodium Infiltration

No Constituent Migration + Sodium Infiltration

Melting Temperature + No constient migration

Melting Temperature + Constient migration

500

600

700

800

900

1000

1100

1200

1300

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Distance from the Fuel Center (r/Ro)

Fuel

Slug

Tem

pera

ture

(o C)

Melting Temperature

radial temperature for U-42TRU-50Zr



● Fission Gas Release (FGR) Prediction

Behavior of intragranular FGR : diffusion theory by Booth

Behavior of grain boundary : multiple bubble distribution model by Hwang

Percentage gas releases according to burnup variation by MACSIS-H

Largely increases at around 1 to 2 at% burnup

Fission gas release at a burnup of 10at% : about 70~80 %.

The predictions by MACSIS-H with the semi-theoretical models agree comparatively well with the experim

ental results from ANL

13

Fission Gas Release and He Release Rate Insertion

Fission gas release data by PahlCalculated fission gas release

0

10

20

30

40

50

60

70

80

90

100

0 5 10 15 20 25 30 35

Burnup (at%)

Fiss

ion

Gas

Rel

ease

(%)

FGR for U-40TRU-50ZrFGR for U-19Pu-10ZrU-20TRU-14.6Zr


● He generation rates insertion

Assumption

- Helium production rates from 6-40 wt% 241Am : 50 ml He per gram of transmuted americium by Meyer

- 241Am weights and the He generation rates : calculated by fuel design spec

14

241Am weights and the He generation rates

Inserting the He generation rates into the code

In the MACSIS-H

- the volume of fission gas generated are recalculated including the He generation rate

Fuel type Content (wt%)241Am weight

(g)He generation rate

U-42TRU-50Zr 2.66 9.1 40.0 ml/165day

U-20TRU-14.6Zr 0.1 0.5 9.36 ml/165day

* time required to achieve 50% transmutation of 241Am : 2.5 years by Walker

Fission Gas Release and He Release Rate Insertion


● Strain limits analysis Cladding strain comparison with He effects as a functio

n of the plenum-to-fuel ratio for HYPER fuel

- The effects on the strain with different plenum sizes :analyzed by the MACSIS-H code

cladding strain by the plenum pressure stress alone

He effects will be a very important factor

- the burnup limit for the plenum-to-fuel ratio of 1.75

: 33at% by the thermal creep strain limit of 1%

- major deformation mechanism

: thermal creep strain in the metallic fuel

- the thermal creep strain of 1%

: used for the burnup limit criteria for metallic fuel

: swelling and irradiation creep of HT9 : very small

1.5 and 1.75 times of the plenum-to-fuel ratios : conservative for satisfying the discharge burnup goal

15

Cladding strain of HYPER fuel according to plenum-to-fuel ratio

plenum-to-fuel ratio

thermal creep strain

Without He

Thermal creep strain

With He

1.50 0.14% at 25at%

0.4% at 25at%

1.75 0.09% at 25at%

0.19% at 25at%

Strain limits analysis

0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

0 5 10 15 20 25 30

burnup(at%)

strain

(%)

1.75 plenum with He 1.75 plenum without He 1.50 plenum with He 1.50 plenum without He


Cladding strain comparison with He effects as a function of the plenum-to-fuel ratio for the critical system fuel

- the HT9 cladding is not conservative for satisfying the discharge burnup goal

: because of the high coolant outlet temperature

- In the metallic fuel of the critical system,

: replacement of cladding material with higher thermal creep resistance may be needed

16

0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

0 5 10 15

burnup(at%)

strain

(%)

1.75 plenum with He 1.75 plenum without He 1.50 plenum with He 1.50 plenum without He

plenum-to-fuel ratio


Without He


With He

1.50 2.27% at 9.8at%

4.8% at 9.8at%

1.75 0.64% at 9.8at%

1.2% at 9.8at%

Burnup limits analysis

Cladding strain of critical system fuel according to plenum-to-fuel ratio


Two kinds of specific design limits

strain limit approach

strongly dependent on temperature and strain rate

cumulative damage fraction (CDF) method

based on linear summation of creep damage

appropriate for non-stationary stress and temperature loadin

g conditions

Weibull analysis

probabilistic method of analyzing life-test experiments

Useful for the evaluation of component reliability

17

Probabilistic estimation of the CDF (Cumulative Damage Fraction)

Derivation of failure distribution functions

Evaluation of

CDF of the X447 fuel pins

Estimation of fuel pin performances

Weibull analysis

MACSIS-H code

HYPER and critical system condition

Calculation Scheme for the steady-state conditions

Calculation Scheme for the transient conditions

Evaluation of

CDF of WPF test fuel pin FCTT failure correlation

Evaluation of cladding performance

WPF transient test data

Derivation of failure distribution functions

WPF test condition

Weibull analysis

Estimate offuel pin performances

t

r Tt

dtCDF

0 ,


limit on the fuel pin failure rate : less than 0.01%

the CDF limit of 0.001 was reasonable

Failure probability of the HYPER fuel pin during steady-state condition

fuel pin failure rates for the 1.5 and 1.75 plenum-to-fuel ratios : 0.017 and 0.003%:

1.75 times of the plenum-to-fuel ratio : conservative by CDF limit

Failure probability of the HYPER fuel pin during transient condition: lower than that of the WPF pin

Because of a higher plenum-fuel volume ratio and lower cladding inner radius vs. thickness ratio

18

Fuel pin performance by Weibull distribution

0.000001

0.00001

0.0001

0.001

0.01

0.1

1

10

0.00001 0.0001 0.001 0.01 0.1 1 10

CDF

Fai

lure

Pro

babi

lity

Weibull distribution at steady state

Weibull distribution at transient state

steady, 10at%, 1.75plenum(critical)steady, 25at%, 1.75plenum(HYPER)

Steady-state failure rate limit

Preliminary CDF limit

steady, 25at%, 1.5plenum(HYPER)

transient, 25at%,700oC,60min (HYPER)

transient, 3at%820oC,67min (HYPER)

transient, 3at%820oC,67min (WPF)

Probabilistic estimation of the CDF (Cumulative Damage Fraction)


19

Performance analysis for the transmutation fuel by MACSIS-H code

Margin to the slug melting temperature

Sufficient margin for the fuel of HYPER and critical system

Eutectic melting limit

One of the main issue, but analysis model is now developing

Detailed analysis needed for the fuel of critical system, because of significant amount of Zr depletion in the middle zone by the constituent migration

Cladding Strain limit

The He effects will be a important factor

Replacement of cladding material may be needed for the fuel of critical system

CDF limit

1.75 times of the plenum-to-fuel ratio was conservative for the fuel of HYPER at steady-state condition

Failure probability of the HYPER fuel pin was low at transient condition

Some experimental test are needed for clarifying the uncertainties of the fuel modeling

Conclusion

korea atomic energy research institute 2004. 11.09-11 the preliminary performance analysis of the...

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