Axial distribution of cesium in heterogeneous FBR fuel pins

Journal of Nuclear

Materials

201 (1993) 46-53

North-Holland

Axial distribution

of cesium in heterogeneous

H. Furuya ‘I, S. Ukai “. S. Shikakura “, Y. Tsuchiuchi

‘The axial migration

behavior

different

configurations

measured

nondestructively

good agreement

between

blanket

position

and analytical

The radial distribution role

in

the

procedure

heterogeneous

fuel pin

Axially heterogeneous fuel pins with different configuration of the internal blanket were irradiated as AHC-1 test in JOYO, to investigate the ccsium bchavior in fuel pins. Fabrication parameters of these pins 0022-3115/93/$06.00

0

1993

Elsevier

The

by an electron

axial

(U.

Pu)O,

distribution

probe microanalyzer.

by applying

fuel pins with of cesium wai A reasonably

a model based on a two-stage

by assuming a rapid migration

process not only across the pellet-pellet

down temperature interfaces.

hut also

axial migration.

The concept of axially heterogeneous fuel is one of the promising design types for the demonstration FBR in Japan. From a viewpoint of fuel pin behavior and integrity toward high burnup, it is inevitable to evaluate the performance of the internal blanket, which provides a cooler temperature region within the active core of an axially heterogeneous fuel pin. Cesium. an abundant and mobile/reactive fission product, migrates axially to lower temperature regions and reacts with blanket fuel to form low density compounds. which cause local pellet swelling. PNC conducted an irradiation test with axially heterogeneous fuel pins with various configurations of an internal blanket in the Japanese experimental fast reactor, JOYO. Results of the postirradiation examination were evaluated with emphasis on the axial migration of cesium, using prcdictions by the mechanistic computer code on cesium migration, MINERVA, developed by PNC [1,2].

2. I. Axially

- 70 GWd/t.

could be explained

1. Introduction

2. Experimental

to

results was obtained

within the solid fuel. The evaporation-condensation

across cracks played an important

“ and K. Idemitsu ”

in two kinds of axially heterogeneous

irradiated

by a y-ray intensity scanning and destructively the experimental

process proposed by Adamson. gradient

of cesium was investigated

of internal

FBR fuel pins

Science Publishers

are shown in table 1. Axially heterogeneous fuel pins of 6.5 mm outer diameter involve Urania-plutonia mixed oxide fuel with an O/M ratio of 1.99 and UO, internal blanket with an O/U ratio of 2.01. The configuration of the internal blanket within the active core region of 550 mm in length consists of four types. Postirradiation examination of two types of fuel pins. AHCO6 and AHCO7, has been completed up to now. Fig. I show the structure of these two types of fuel pins. The two internal blankets of 60 mm length are placed in the AHC06 pin at an axial distance of 100 mm and 340 mm from the bottom of the active core. In the AHC07 pin, they are positioned at a distance of 40 mm and 400 mm from the bottom. Other test pins now continue to be irradiated in the JOY0 AHC-2 test. 2.2.

Irrudintion

condition

The axially heterogeneous fuel pins wcrc inserted into the AHC-1 test rig, which is shown in fig. 2. These pins were installed in the center of the rig and they wcrc surrounded with standard driver fuel pins. Irradiation was conducted at the position of 2B2 in the JOY0 core from the 14th through the 19th cycle. The peak burnup attained was 72100 MWd/MtM and 71200 MWd/MtM for the AHC06 and AHC07 pins. respcctively. The peak of the fast neutron fluence reached about 9.7 x 10’” n/m* (E > 0.1 MeV) for both fuel pins. The peak linear heat rate of the core fuel was about 45 kW/m at the beginning of life. while the one

B.V. All rights reserved

H. Furuya et al. /Axial distributionof Cs in FBR fuel pins Table 1 Fuel pin fabrication parameters 6.5 mm 5.65 mm

Cladding outer diameter Cladding internal diameter Mixed oxide pellet Pellet outer diameter Pellet density O/M ratio PUO, /(UO, + PUO,) Internal blanket Pellet outer diameter Pellet density O/U ratio 235Ucontent

5.41 mm 92.52% TD 1.99 17.64 wt%

(4

Loa Iding Position in JOY0

5.41 mm 93.81% TD 2.01 0.22% wt%

of the internal blanket increased to almost 10 kW/m at the end of irradiation due to the breeding of 239Pu from 238U. AHCOG pin7

rAHC07

/

2.3. Postirradiation examination

:,:

:s

Cut of AHC Rig

(c) Position of Test Pins

Fig. 2. Configuration of AHC-1 test rig and loading position of the rig in JOYO.

After the non-destructive examinations, both fuel pins were cut, and ceramographic samples were prepared in a metallography cell of the FMF. Electron

210

220

160 100

/

/

After a sufficient radioactive decay period, the irradiated AHC-1 test rig was transfered into the Fuel Monitoring Facility (FMF) which is directly connected with the JOY0 reactor. The test rig was dismantled of fuel pins, and a series of nondestructive examination was carried out for the axially heterogeneous AHC06 and AHC07 pins, including the fuel column length measurement, pin diameter measurement and fission gas release measurement. The y-ray spectra of 137Cs, 134Cs, 94Zr and l”Rh were taken at intervals of 1 mm in length along the active fuel column.

::: :.. 1

(b)/Radial

pin

/$$I

Eai

330

:.

180

300

II 0 AHCOG pin

150

~~~

k!jj

Core Bottom

93

1’1 i AHC07

Core Top

q Internal

Blanket

Fig. 1. Structure of axially heterogeneous fuel pins.

pin

Axial ~jsrr~6urjon of C’s in Fi3R ,fuelpins

H. Furzqa et al. /

4x

AHC07 pin

AHCOG pin 128

77

220

363

Axial

Position

Fig. .7. Axial profile

506

0 77

649

(mm)

of “‘Cs y-ray intensities of Al-K.06

Fig. in the number bottom

in fuel pins

3 shows axial profiles of “‘CS y-ray intensities AHCO6 and AHCO7 pins. In this figure, the on the abscissa indicates the distance from the of the fuel pins. The active fuel column covers

(mm)

AHC07 pin 1.02

H

0.76’

: internalBlanket

F----I

and Thermal Insulator +

-

b---i

(?

: InternalBlanket

and Thermal Insulator

0.76.

B

d

J

01 77

220 Axial

363 Position

506 (mm)

649

and AHCO7 pins.

AHCOG pin 1.02

,s

Position

506

a region from 90 through 640 mm. Although a distinct cesium peak is not found at the interface between core and internal blanket fuel peilets, a series of small spikes can be clearly seen at the positions corresponding to the pellet-pellet interface in the core fuel region. In the internal blanket region, inhomogeneous distribution of ccsium can be slightly observed. In order to discern this behavior more clearly, axial profiles of the intensity ratio of ‘37Cs to the immovable element “‘Zr were taken as shown in fig. 4. in which ?Er is one kind of indicator for the amount of cesium generated by fission. It can be recognized that this ratio becomes relatively high just at the interface be-

3. Results

of cesium

363

Axial

probe microanalysis was conducted also to characterize the location and chemical form of accumulated cesium in the fuel pins.

3.1. Axial profiles

220

I

649

t---i

01

77

220 Axial

363

506

Position .

(mm)

Fig. 4. Axial profile of 13’Cs/ ““Zr y-ray intensity ratio of AHC06 and AHC07 pins.

H. Furuya et al. / Axiui d~t~b~t~un

core and internal blanket fuel pellets. These findings suggest that cesium migrated axially to extent of length of a pellet, and slight accumulation of cesium took place at the interface between core and internal blanket fuel pellets.

tween

Four and three samples were prepared from the fuel column of each of the AHC06 and AHC07 pins in the longitudinal and transverse cross section, respectively. Axial locations of four longitudinal samples covered the interfaces of core fuel and internal blanket. In core pellets, a remarkable fuel restructuring was observed, and the central hole, and columnar and equiaxial grains were clearly distinguished. On the other hand, the fabrication structure remained in internal blanket pellets with only radial and longitudinal cracks. Fig. 5 shows the microfine features around the location

.!)--“iIn

Upper Internal

of Cs in FBR fuelpins

49

of the upper interface between the core and internal blanket fuel pellets in the AHC06 pin. In this figure, radial and longitudinal profiles of cesium intensity measured by the electron probe microanalyzer are also shown, together with the fine structure of the blanket pellet. Cesium intensity is higher at the peripheral region of the blanket pellet as indicated by the radial distribution curve of cesium along the direction of A-A’. The longitudinal profile along direction B-B’ indicates that cesium intensity in the internal blanket region decreases with increasing distance from interface between core fuel and internal blanket. This result is in good agreement with the y-ray intensity measurement. It can be also clarified that cesium accumulates in the cracks and voids of the internal blanket. In order to obtain a more detailed distribution of cesium at the interface between core and internal blanket fuel pellets, the X-ray map of cesium at position of C is shown. A quite similar behavior was observed for the AHCO? pin.

Blanket

Core

Fig. 5 Ceramography and results of electron probe microanalysis at the upper internal blanket and core fuel of AHCOB pin.

50

H Furuyu

et

al. / Axial distribution

4.Discussion 4.1. Mechanistic model of cesium migration; ERVA code

the MIN-

From a number of studies on axial cesium migration in mixed oxide fuel pins, [3], it has been found as shown in this study that cesium migrates down the temperature gradient, resulting in accumulation of cesium and subsequent swelling of cesium compounds at the interface between core and blanket fuel pellets. Bases on results of out-of-pile experiments and detailed inspection of postirradiation examination data, Adamson et al. [4-71 have proposed a model that describes transport of chemically combined cesium as a two-stage process, one stage being dominant within the solid fuel and the other at the pellet-pellet interfaces. The cesium migrates relatively rapidly within the solid, possibly as a result of or in conjunction with oxygen thermomigration. Although this mechanism have not been well understood, the observed concentration of cesium within solid pellets in irradiated fuel pin can be given approximately by the following Arrhenius-type equation: N(r)

=A exp(-Q*/RT(r)),

(1)

where N(r) is the cesium concentration at radial position r, and T(r) is the corresponding temperature at Y. * is the effective heat of transport of cesium. The Q transport of cesium across the pellet-pellet interfaces occurs as a result of an evaporation-condensation process, and the total flux of cesium, 4, across the interface can be expressed by the following equation: d, =k[N(h)

-N(C)]

exp( -AH,/RT(r)),

of Csin FBR fuel pins

+1 lnitical

Increase

Condition

(t=O)

b of Time

Step

t=t+_ltB 1 Calculate

Fuel Temperature

Calculate Product

amount

t of Fission

. Calculate Radial Distribution

Cesium

1 Calculate O/M Ratio Oxygen Potentical

and

1 Calculate

Chemical

Equilibrium

Calculate Cesium Axial Transport by Vaporization-Condensation Mechanism

I

Calculate Fuel Swelling Cs-Fuel Reaction

due to

(2)

where N(h) and N(c) are the concentration of cesium on the hotter and cooler side at the pellet-pellet interface, respectively. k is the mobility of cesium in the gap, and AH, the partial molar heat of vaporization of cesium. In this study, a mechanistic computer code, MINERVA (h4igratiog Behavior of Volatile Cesium &alysis) was developed to predict cesium migration and reaction swelling in the irradiated mixed oxide fuel in PNC, by incorporating some modifications in thermodynamic treatment into the above-mentioned model based on the two-stage process [1,2]. In connection with the radial distribution of cesium, lmoto [8] proposed another analytical model, but eq. (1) was applied here to make the analysis simpler. Fig. 6 shows the main flow chart of the MINERVA code. The analytical system is two dimensional in R-Z directions, and the

Fig. 6. Main computational flow of MlNERVA code.

radial distribution of cesium was initially computed from the fuel temperature profiles and the amount of cesium generated in the fuel pin. Oxygen redistribution is calculated according to Aitken’s model [9]. The relationship between O/M ratio and oxygen potential was determined, based on Woodley’s model [lo] in the region less than 2.00 and on Blackburn’s model [ll] in the region higher than 2.00. Amounts of stable cesium compounds formed at the pellet periphery are computed from both results of calculation of O/M ratio and oxygen potential, and then applied to the calculation of fuel swelling. Itera-

H. Furuya et al. / Axial distributionof Cs in FBR fuel pins

s

Core Center

2

_

5; 0 z : ‘q

-

until end of irradiation. This code was verified using the homogeneous fuel pins irradiated in JOY0 and Phenix [2].

of AHCOG Pin

- Experimental _-_ 57kJ/mol

Data

4.2. Radial distribution of cesium

47kJ/mol

l.O-

I ‘; : : 0

51

.

OO

MINERVA code was applied to the analysis of the cesium migration behavior in AHC06 and AHC07 pins. As already described above, cesium radial distribution in the fuel pellet can be calculated by eq. (1). Radial distribution of cesium in the fuel was measured by an electron probe microanalyzer. Fig. 7 shows the measured concentration of cesium across the fuel pellet at the axial mid-plane of the AHC06 pin. The value of 57 kJ/mol was obtained for Q*, from the best correlation of experimental results with eq. (1). The profile, using the value of 47 kJ/mol, which is obtained from the homogeneous fuel pin, is also shown for the purpose of comparison. For both of homogeneous and heterogeneous fuel pins, it was recognized that the MINERVA code can give reasonable distribution of cesium across the fuel pellet using a single value of Q*.

*

0.8

0.6

0.4

1.0

r/r0 Relative

Distance

from

Pellet

Center

of radial cesium distribution between measured results by electron probe microanalyser and MINERVA code prediction.

Fig. 7. Comparison

tive calculation was continued until the calculated oxygen level showed agreement with the assumed one within a reasonable limit. Based on the calculational results of radial and axial distributions of cesiumuranoplutonate, vapor transport of cesium due to evaporation-condensation processes is calculated. Concomitant reaction swelling can also be calculated. These calculations are carried out at intervals of a time step

4.3. Axial migration of cesium The MINERVA code was applied to the analysis of the axial concentration profile in two axially heteroge-

AHCOG pin

AHC07 pin

951

I

O5Y!7riEe

750

g51

0 50

Axial Position (mm)

0’

*

.

.

.

Axial Position

.

225

400

575

750

Axial Position (mm)

1

oAxial Position

Fig. 8. Results of axial cesium distribution predicted by MINERVA code for AHC06 and AHC07 pins.

and AHC07 fuel pins. 177.7 kJ/mol was used as AH,, the partial molar heat of vaporization of ccsium, as recommended by Adamson [7]. At first, a reasonably good agreement between the observed and analytical results of the axial cesium profile was not obtained. There are many cracks perpendicular to the axial direction of the fuel pellet as shown in the longitudinal ceramography of fig. 5. Then, it is likely that vapor transport of cesium takes place by an evaporation-condensation process not only across pellet-pellet interfaces but also across such cracks. Ceramography shows that core and blanket pellets were longitudinally cracked into eight and three pieces, respectively. These cracks were treated in the same manner as the pelletpellet interface for the cesium vapor transport by the evaporation-codensation process. The mobility of ccsium in the gap was determined by applying the MINERVA code to the analysis of marked peaks at both the upper and lower interfaces between the core and blanket fuel pellets in the homogeneous fuel pins. Fig. 8 shows the results of axial migration of cesium in AHC06 and AHC07 pins calculated by the MINERVA code using the parameter mentioned above and the exact irradiation history and fuel specification. These results show good agreement with the expcrimental ones illustrated in fig. 3. It suggests that cesium can migrate only to the extent of the length of a pellet. The major parameters influencing cesium axial migration should be the operating power level, the fuel burnup and the O/M ratio. In case of this irradiation test, the axial decrease of the tempcraturc gradient at the pellet periphery within the shorter fuel column seems to contribute to retardation of cesium axial migration. In this model, the difference of radial ccsium transport through the solid phase and through the gas phase can not be taken into consideration. Therefore. the small peaks at the interfaces between the core pellets or radial cracks may result from the unbalance bctween the cesium flows to and from the gap, because the radial transport of cesium easily takes place through the gas phase in the gap.

neous AHC06

4.4. Effect

of fuel-cladding

chemical

interuction

It was assumed in the MINERVA code that thermochemical stability of cesium compounds is in a SCquence of CsI, Cs,Te, Cs,MoO,, Cs,UO, and Cs,UO, as a function of cesium potential. The chemical reaction of cesium with cladding elements can not bc considered in the present model. The most dominant

~~~

2Cs+Cr+202-Cs2Cr04 2

2cs+

_

-

: cr*os+

2 O~-Cs&ro~

CszCr04 Cs~UO,(lOOOK)

I

-4 -32

-31

-30 log

Fig. 9. Equilibrium

I

I

-29

-28

(PoJatm)

oxygen and cesium potentials

and Cs.UO,

-

for (‘a T<‘rO,

formations.

reaction product is Cs,CrO,, and the possibility ot formation of this compound was investigated when the gap between fuel and cladding is open. Fig. 9 shows results of the calculation of the oxygen and cesium pressure for both formation of Cs2Cr0, in which Cs,UO, formation is fixed and Cs,UO,, around 1000 K corresponding to the temperature at the pellet periphery. Comparing the cesium potential at constant oxygen potential, it is recognized that the Cs,CrO, compound can be formed at the inside of the cladding at temperatures less than 900 K as also pointed out by Fee an Johnson [12]. Also under the condition of a closed gap, the reaction of cesium with the cladding is expected to occurc by the CCCT mechanism proposed by Adamson [ 131. This part of ccsium consumed by a reaction with the cladding might bc expectccl to contribute to uniform axial distribution, and therefore should be taken into account in future analysis.

5. Conclusion Two types of axially heterogeneous fuel pins with different configurations of the internal blanket position were irradiated in JOYO. It was found from the y-ray spectrum and the electron probe microanalysis that axial migration of cesium occurred to the extent of the distance of a pellet in length, and slight accumulation of cesium took place at the interfaces between core and internal blanket fuel pellets. These experimental results were evaluated using the mechanistic computer

H. Furuya et al. / Axial distribution of Cs in FBR fuel pins

code MINERVA, which is basically in accordance with Adamson’s and Vaidyanathan’s model. Applying the same parameters as used for usual homogeneous fuel pins, a reasonably good agreement between the experimental and analytical cesium distributions could be obtained for irradiated axial heterogeneous fuel pins. Based on the model of evaporation-condensation processes, it was demonstrated that Cs-vapor transport plays an important role in axial cesium migration not only across the pellet-pellet interfaces but also in the fuel cracks. It was also suggested that reaction of cesium with cladding elements should be taken into account to predict accurately the cesium axial migration behavior.

References [l] M. Ishida et al., Proc. Fall Meeting of the Atomic Energy Society of Japan, 1987,577.

53

[2] Y. Saito et al., Proc. Fall Meeting of the Atomic Energy Society of Japan, 1988, H14. [3] For example, R.A. Karnesky, J.W. Jost and 1.2. Stone, Proc. Int. Conf. on Fast Breeder Reactor Fuel Performance, March 5-8, 1979, Monterey, CA, p. 343. [4] M.G. Adamson, Proc. IAEA Specialists’ Meeting on Fuel-Cladding Interaction, Tokyo, IWGFR/16 (1977) p. 108. [5] M.G. Adamson and E.A. Aitken, Trans. Am. Nucl. Sot. 17 (1973) 195. [6] M.G. Adamson and S. Vaidyanathan, Trans. Am. Nucl. Sot. 38 (1981) 289. [7] S. Vaidyanathan and M.G. Adamson, Trans. Am. Nucl. Sot. 38 (1981) 291. [8] S. Imoto, J. Nucl. Mater. 169 (1989) 336. [9] E.A. Aitken, J. Nucl. Mater. 30 (1969) 62. [lo] R.E. Woodley, J. Nucl. Mater. 96 (1981) 5. [ll] P.E. Blackburn, J. Nucl. Mater. 46 (1973) 244. 1121D.C. Fee and C.E. Johnson, J. Nucl. Mater. 96 (1981) 80. [13] M.G. Adamson and E.A. Aitken, Proc. Int. Conf. on Fast Breeder Reactor Fuel Performance, March 5-8, 1979, Monterey, CA, p. 536.