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ANL-7602 Reac to r Technology

ARGONNE NATIONAL LABORATORY 9700 South C a s s Avenue Argonne, I l l inois 60439

POSTIRRADIATION EXAMINATION O F U - P u - Z r FUEL ELEMENTS IRRADIATED IN EBR-I1

TO 4 .5 ATOMIC PERCENT BURNUP

W . F. Murphy, W . N . Beck, F. L. Brown, B. J . Koprowski, and L. A. Ne imark

Metallurgy Division

November 1969

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

T A B L E O F C O N T E N T S

Page

7 ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

7 I. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . o . . o a

11. FUEL-PIN MATERIALS. . I . . . . . . . . . , e . e . e I 8

111. CLADDING MATERIALS. - . . . . * . . . . . . . * . - . . D o 12

IV. DESIGN O F THE FUEL ELEMENTS. a . . a . a e e e e e 14

V . IRRADIATION IN EBR-I1 . a . . . . . . . * . . . . . . . . . , . 16

VI. POSTIRRADIATION EXAMINATION . e . . . . . . . e e e e 2 1

A.

B.

C.

D.

E.

F.

G.

H.

I .

J.

K.

L.

Di sas sembly of the Capsules . . . . . . . e . e . . e 21

Dimensions of the Fuel Elements . . . . a . , , a a . . 21

Weights and Volumes of the Fuel E l e m e n t s . . . . e . . . 23

Dimensions of the Fuel P ins . . . . . . e . . . . e e 23

Axial G a m m a Pro f i l e . . e . . . . . e . . I . . . . a 2 7

Determinat ion of Sodium L e v e l . . . . . . . . . e . . . . . 2 7

Analysis of F i s s i o n G a s . . . . e . . . e . . . . a . e . 27

Metallographic Examination of the F u e l . . . . , e . e 2 8

Examination of Separa ted Annular Zones . . . . . . e . e . 29

Burnup Analyses . . . . . . I . * . . . . . . . . . e a - . I 38

Metallography and Microprobe Analysis of the Cladding . 3 8

Hardness T e s t s on Cladding. . . . . . . . . . . . . . e . e . . e 46

. . e e e

VII. DISCUSSION.. * . * . . . * . . . . . . . . . . . . . m . o . . e 46

A. Dimensional Changes of the Fue l E l e m e n t s . e . a . 46

B. Fuel Swelling and Fiss ion-gas Release . a . . , e e e I 48

C, Fuel-pin Behavior . * . . . . . . a . . . e . . . . 50

D. Fuel S t r u c t u r e . . - . e . . . . . . , * . . . . e . 51

E. Fiss ion-product Distr ibut ion and Composition Change. , 55

A

T A B L E O F C O N T E N T S

Page

F. B u r n u p A n a l y s e s . o , ,, I 57

G. Cladding and Fuel-Cladding In te rac t ion . /I , ~ 57

VIIIa CONCLUSIONS. e ,, , 58

ACKNOWLEDGMENTS . ,, 54.

60 REFERENCES, e a n O o , j D , j D o D D y _j , , , , . "

Y

LIST O F FIGURES

Tit le No. Page

1. Inject ion-cast U-14Pu-12Zr Alloy--Batch R250 . I . . a . . 10

2. Inject ion-cast U-15Pu-9Zr Alloy--Batch R253Z e . a 11

3. Schemat ic of Fuel Element with Res t r a ine r . I e e 14

4. Axial T e m p e r a t u r e Distr ibut ion in Fuel Element ND35. e 20

5. Radial T e m p e r a t u r e Distribution in Fuel Element ND35. e 20

6. 24

7 , Axial Distribution of Separat ions in Fuel Columns. , e 25

8. 28

9. Axial Development of Annular Zones in Fuel P i n ND35. a ~ 30

10. End Effects i n Fuel Co lumns . . . e a a e 31

11. Separat ions i n Fue l Columns, As-pol i shed . a Ij 32

12. Variat ions in Poros i ty of U-15Pu-9Zr Fuel , As-pol ished . a , 33

13. In te rna l Cracking in U-15Pu-9Zr Fuel , As-polished ~ e a . . 34

14. Atypical Mic ros t ruc tu res in U - P u - Z r Fuel . . e a a .1 35

15. Swir led Effect near Bottom of Element ND35 . e . e 35

16.

G a m m a Scan and Neutron Radiograph of Fuel Column ND25

Effect of Swelling in Metall ic Fuels on F iss ion-gas Re lease

Banding at Boundary between Middle and Center Zones of ElementND35. D . . a . * . " . I . .. e e a * , I) 36

17. Cel lular Mic ros t ruc tu re in Element ND35 a a e e e e a 36

18. Separa ted Annular Zones of Element ND30. . . e , e 37

19. Compositional Variat ions i n Annular Zones of U - P u - Z r Fuel P i n s . . . - . . - . . . * . e - . . . * . . . . a - . , - . 38

20. Fuel-Cladding In te rac t ions , As-polished . . . . (I e 40

21. Cladding C r a c k s . . * . e . e . . . - e 41

22. S t ruc tu res of I r r ad ia t ed Claddings . , . . e e e 42

* 23. S t ruc tu res of Heat - t rea ted Claddings . . e . . . . a 43

24. Typical Distr ibut ion of Elements in Surface Laye r i n I r ra - diated Type 316 Stainless Steel Cladding . j. e a . e 45

25. AV of Fuel Element v s C r o s s Section of Cladding a e 48

26, Length Change vs Tempera tu re of F u e l . a . e . a e e 51

27. 55 Compositions of the Annular Zones in the U - P u - Z r Sys tem.

5

LIST O F TABLES

No. Ti t le Page - I . Fuel -e lement Identification . . . . . . . . . . . . . . . . . . . . . . 9

I1 . P r e i r r a d i a t i o n Fuel-pin Data . . . . . . . . . . . . . . . . . . . . . 12

I11 . Chemical Compositions of Claddings . . . . . . . . . . . . . . . . 13

IV . Measured Dimensions of Tubing fo r Cladding . . . . . . . . . . 13

V . Design F a c t o r s of Fuel Elements . . . . . . . . . . . . . . . . . . . 15

VI. P r e i r r a d i a t i o n Measuremen t s of Fuel Elements . . . . . . . . . 16

VI1 . F i s s i o n Ra te s in EBR-I1 Core Location 4D3 at 45 MW . . . . . 17

VI11 . Subassembly Posi t ions and Calculated Burnups . . . . . . . . . 18

IX . Hea t -gene ra t ionDa ta . . . . . . . . . . . . . . . . . . . . . . . . . . 18

X . Init ial Fue l and Cladding Tempera tu res . . . . . . . . . . . . . . 19

XI. Calculated End of I r rad ia t ion P r e s s u r e and S t r e s s Data . . . 2 1

XI1 . Dimensional Changes of the I r r ad ia t ed Fue l Elements . . . . . 2 2

XI11 . Weight and Volume Changes of I r r ad ia t ed Fue l Elements . . 23

XIV . Length Changes of 1rradi.ated Fuel Columns . . . . . . . . . . . 25

XV . Calculated Volume I n c r e a s e s of I r r ad ia t ed Fuels . . . . . . . . 26

XVI . Fis s ion Gases (Xe t K r ) and Hel ium Collected f r o m P lenums of I r r ad ia t ed Fuel Elements . . . . . . . . . . . . . . . . 28

XVII . Densi t ies of Annular Fuel Sections . . . . . . . . . . . . . . . . . 37

. . . . . . . . . . . . . . . . . . . . XVIII Burnups by Technet ium Analysis 38

POSTIRRADIATION EXAMINATION O F U - P u - Z r F U E L ELEMENTS IRRADIATED IN EBR-I

TO 4.5 ATOMIC PERCENT BURNUP c

W. F. Murphy, W. N. Beck, F. L. Brown, B. J . Koprowski, and L. A. Ne imark

ABSTRACT

Sixteen U-15 w/o PU-10 w/o Z r meta l l ic fuel e le - men t s w e r e i r r ad ia t ed to 4.5 a / o burnup at 10 kW/ftwithout fa i lure . Cladding m a t e r i a l s w e r e Types 304L and 316 s ta in- l e s s s t ee l , Hastel loy-XI and Hastelloy-X-280. The maxi- m u m cladding t empera tu res ranged f r o m 605 to 650°C. Diamet ra l and axial expansion of the cladding w e r e both l e s s than 0.3'-%. The fuel had swelled to the cladding, and 14 of the fuel columns elongated and two dec reased in length . The in te r face between the fuel and the cladding showed evidence of a react ion between the two. The smallest reac t ion l a y e r (-1 mil) was i n the Type 316 s ta in less s tee l .

Annular zones had developed in the fuel. These zones a r e principally t empera tu re dependent and a r e probably assoc ia ted with phase t ransformat ions in the fuel, F i s s ion products and m a j o r fuel consti tuents w e r e found to have mig ra t ed between the t h r e e zones.

I. INTRODUCTION

Sixteen exper imenta l meta l l ic fuel e lements w e r e i r r ad ia t ed in These exper iments a re EBR-I1 to 4.5 a /o burnup without any fa i lures .

p a r t of a s e r i e s of i r rad ia t ion s tudies to demons t r a t e the per formance of meta l -a l loy fuel e lements at r e a c t o r conditions necessa ry f o r the economic acceptance of such fuels for fast r e a c t o r s e rv i ce . The goal is to achieve, through carefu l analytical in te rpre ta t ion of the r e su l t s obtained, sufficient da t a to des ign rel iably and to pred ic t safely the per formance of metal- alloy fuel e lements for commerc ia l LMFBR's .

The U - P u - Z r fuel s y s t e m was or iginal ly investigated in F r a n c e as a potential fuel for fast r e a c t o r s . l V 2 pe r fo rmed in the USA, prompted by the higher solidus t e m p e r a t u r e (>11OO"C)

F u r t h e r labora tory s t u d i e ~ ~ ' ~ w e r e

compared with the solidus t e m p e r a t u r e of the U-Pu-Fz* alloys ((900°C) that w e r e then being investigated. and Type 304 s t a in l e s s s t ee l cladding i n out-of-pile t e s t s at t empera tu res up to 8OOOC was be t t e r than had been e x p e ~ t e d , ~ , ~ ~ based on the phase r e - lationships among the me ta l l i c components. A reac t ion between oxygen in the s ta in less s t ee l and z i rconium i n the fuel is believed to have resu l ted i n the formation of a b a r r i e r l aye r that r e t a rded interdiffusion.

Compatibility between the U - P u - Z r fuel

About the s a m e t i m e that these l abora to ry t e s t s w e r e being con- ducted, exper iments indicated that the r a t e of swelling of a meta l l ic fuel d e c r e a s e d when the fuel was allowed to expand f r o m 25 to 30 v / o and thereby allow apprec iab le f iss ion-gas releas e. l1 A fuel- e lement design that exploited this effect was devised; it incorpora ted a sodium-fil led annulus to p e r m i t 30% rad ia l swelling and a plenum above the fuel pin with sufficient volume to accommodate the r e l eased f i ss ion gas at low p r e s - sure ,12 l ished that higher burnups without fuel- e lement fa i lure could be attained with a U - P u - Z r alloy than with a similar U-Pu-Fz alloy at comparable t empera tu res and i n s imi l a r ly designed e lements . l1

Subsequently, i r rad ia t ions in a thermal r e a c t o r (CP-5 ) had estab-

The va r i ab le s of fuel composition, fuel- e lement-geometry, and

C o m m e r - cladding w e r e combined in the design of an exper iment to test fuel e lements with a l a r g e gas plenum t o burnups of 10 a/o. cially available s t a in l e s s s t ee l s and nickel-base alloys w e r e se lec ted €or claddings of t hese high-burnup e lements because of the h igh- tempera ture s t rengths of t hese materials. The e lements w e r e to be i r r ad ia t ed in a fast r e a c t o r (EBR-11) to de t e rmine the potential of this des ign concept. The exper iments w e r e organized to obtain information applicable to the des ign of an LMFBR fuel e lement .

Sixteen encapsulated fuel e lements w e r e fabr ic at ed , i r r ad ia t ed to 4.5 a /o burnup, and examined. and complete r e su l t s of the exper iment are r epor t ed h e r e .

P r e l i m i n a r y r e su l t s have been r e p ~ r t e d , ' ~

11. FUEL-PIN MATERIALS

The fuel pins (the fue l alloy within the cladding) w e r e cyl indrical rods nominally 13 in . in length and 0.144 in. i n d i ame te r ; the nominal com- posit ion was U-15 w/o Pu-10 w/o Zr .** Vycor molds , and f la t ends w e r e machined on both ends of the cas t ings .

The rods w e r e injection c a s t into

*Fissium (Fs) is the name given the equilibrium mixture of metallic fission products resulting from the pyro- metallurgical reprocessing of uranium-plutonium alloys. A typical composition contains 0.5 w/o Zr, 2.8 w/o Mo, 4.3 W/O Ru, 0.7 W/O Rh, aiid 2.5 w/o Pd. Fizziuin (Fz) differs from fissium (Fs) in that the combined quantity of ruthenium, rhodium, aiid palladium has been reduced from 7.5 to 4 w/o, aiid the amount of zirconium has been increased to "2.5 w/o; molybdenum remains at -2.8 w/o.

**Unless otherwise specified, a l l compositions are stated in weight percent.

a

Code F u e l Designat ion Compo si t ion Type of of Element (W/O) Cladding

The fuel pins were visual ly inspected for sur face defects and X-rayed to de t e rmine cast ing uniformity. w e r e made f o r this experiment .

Two batches of cast ings (R250 and R253Z)

Code F u e l Designat ion Composi t ion Type of

Cladding of Element (W/.)

Compositions of the fuel alloys were de te rmined by chemical analy- s e s of samples taken f r o m each me l t . m a j o r consti tuents w e r e a s follows:

The r e su l t s of the analyses for the

compo s i tio n (w / 0 )

Batch R250 73.4 13.8 1 2 . 2 Batch R253Z 75.9 14.8 9 .4

Each value is the average of two analyses of the s a m e mel t . pins f r o m batch R250 will be designated a s U-14Pu-12Zr and those f r o m batch R253Z as U - 15Pu-9Zr . given in Table I. ing r e su l t s :

In this r e p o r t ,

The fuel composition of each fuel e lement is Analyses of the fuel for impuri ty content gave the follow-

Impuri ty Content (ppm by weight)

0 N H C Si F e A1 - - - - - - - Batch R250 810 36 1.5 32 723 400 50 Batch R253Z 380 11 1 . 7 32 673 300 25

ND28 ND4 1 ND32 ND43 ND25 ND27 ND26 ND29

U- 15Pu-9Zr U- 15Pu-9Zr U- 15Pu-9Zr U- 15Pu-9Zr U- 14Pu- 1 2 Z r U- 14%- 1 2 Z r U- 14Pu- l 2 Z r U- 14Pu- l 2 Z r

304L SS 304L SS 316 SS

Hastel loy-X 304L SS 304L SS 316 SS 316 SS

ND30 ND3 1 ND33 ND34 ND35 ND3 7 ND39 ND44

U- 14Pu- l 2 Z r U- 14Pu- l 2 Z r U- 14Pu- l 2 Z r U- 14%- 12Zr U- 14Pu- l 2 Z r U- 14Pu- l 2 Z r U- 14Pu- l 2 Z r U- 14Pu- l 2 Z r

316 SS 316 SS

Hastel loy- X Hastel loy-X Hastel loy-X

Hastelloy-X- 280 Hastel loy-X- 280 Hastel loy-X- 280

The isotopic composition of the u ran ium and plutonium in both batches was : 238U, 6.81700; 235U, 93.1970; 239Pu, 91.35700; 240Pu, 7.85700; 241Pu, 0.7770; and 242Pu, 0.03570. The 23BU percentage includes small amounts of 234U and 236U.

Metallographic examinations were made of representa t ive samples The photomicrographs ( s e e F igs . 1 and

The white phase is oxygen- f r o m the two batches of cast ings. 2 ) show a r a the r complicated mic ros t ruc tu re . s tabi l ized M, zirconium containing some uranium and plutonium in solid

10

solution. l4 Microprobe analysis indicated that the angular inclusions were high i n z i rconium and s i l icon, and their morphology sugges ts that they are carb ides . phases : nium; the inner phase is high in z i rconium and oxygen. The remainder of the s t ruc tu re often shows evidence of coring and of dendri t ic f o r m s , The phases in equi l ibr ium at room tempera tu re a r e a uran ium, 6(UZr2) , and

The thin, white su r face layer [ s e e Fig. 2(c)] cons is t s of two the outer phase contains principally si l icon, oxygen, and z i rco-

C(U ,Pd .

46526 500X

(a) Casting No. 4, Top, Etched, Ames Reagent

46 522 500X

(b) Casting No. 2, Center, Etched, Ames Reagent

46527 500X

(c) Casting No. 4, Bottom, Etched, Arnes Reagent

Fig. 1

Injection-cast U-14Pu-12Zr Alloy-- Batch a50

11

i

n

FP -92 51 500X

(a) Casting No. 2 , Top, Etched, Ames Reagent

Surface of Casting

4

FP-9249 500X

(b) Casting No. 1, Bottom, Etched, Ames Reagent

Fig. 2

Injection -cast U -15Pu -9Zr Alloy - - Batch R253Z

FP-9254 500X

(c) Casting No. 2 , Bottom, Etched, Ames Reagent

The fuel pins w e r e weighed to the nea res t mi l l i g ram; densi t ies w e r e de te rmined by i m m e r s i o n in bromobenzene. and densi t ies of the fuel pins a r e given in Table 11.

The weights, vo lumes ,

Measurements of the d i ame te r s of the fuel pins w e r e made with a dial m i c r o m e t e r graduated in 0.0001-in. increments . Two sets of readings

12

w e r e taken perpendicular to each o ther at i n t e rva l s of 1 in . along the length of each pin. Twelve measu remen t s of the d i a m e t e r w e r e m a d e on each pin. S ince the su r faces w e r e a s - c a s t , s o m e roughness exis ted; howeverp it did not n e c e s s a r i l y show up on these d i a m e t r a l measu remen t s . w e r e machined, the length of each fuel pin was m e a s u r e d in a f ix ture in which a s t anda rd length was used to z e r o a d ia l ind ica tor . dimensions a r e shown in Tab le 11.

Af ter the ends

The l inea r

TABLE 11. P r e i r r a d i a t i o n Fuel-pin Data

Code Average Average Designation Lengtha Diameterb Weight Volume Density of E lement ( in , ) ( in . ) (8) (cm3) (g / cm3)

ND28 ND41 ND32 ND43 ND25 ND27 ND26 ND29 ND30 ND31 ND33 ND34 ND35 ND37 ND39 ND44

13,000 13.000 13.000 13.000 12.998 12,995 13.001 13.000 12.998 12.998 12.998 12.995 13.000 12 .996 13.001 13.000

0,1436 0.1440 0.1436 0.1437 0.1441 0,1440 0.1436 0.1428 0.1432 0.1444 0.1427 0.1430 0.1443 0.1440 0.1434 0.1444

53.860 53.951 53.890 53.874 52.490 52.462 52.695 51.128 51.290 52.721 51.151 51.264 53.025 52.101 52.217 52.342

3.421 15.75 3.429 15.74 3.422 15.75 3.420 15.75 3.451 15.21 3.432 15.29 3.429 15.37 3.345 15.28 3,353 15.29 3.435 15.35 3.346 15.29 3.355 15.28 3.459 15.33 3.426 15.21 3.413 15.30 3.459 15.13

aAverage of 2 measu remen t s . bAverage of 12 m e a s u r e m e n t s on each pin; to ta l number of

measu remen t s 192. r ange , 0,1415-0.1454 in . ; and 89% of r ange , 0,1420-0.1449 in.

Total r ange , 0,1399-0,1505 i n . ; 96% of

111. CLADDING MATERIALS

Cladding m a t e r i a l s w e r e Types 304L and 316 s t a in l e s s s t ee l , Hastelloy-X, and Hastelloy-X-280 that a r e s t r o n g e r than either of the types of s t a in l e s s steel. t e m p e r a t u r e s , Type 316 s t a in l e s s s t e e l is somewhat s t r o n g e r than Type 304L. th ickness of 20 mils; the wall th icknesses of the o the r s w e r e 15 mils. The ins ide d i a m e t e r (ID) of the Hastelloy-X-280 was 0.176 in . ; the o the r tubes w e r e of 0.166-in. ID. The tube s tock was obtained f r o m commerc ia l s ou rc e s .

The latter two a r e n icke l -base al loys At elevated

The Type 304L s t a in l e s s s t e e l tube stock had a nominal wall

The Type 304L s t a in l e s s s t ee l tube s tock was rece ived as annealed s e a m l e s s tubing with a 0.156 f 0.0005-in. ID and a 0.020 f 0.0005-in. wall thickness . de te rmined by the vendor , was 87,000 ps i , and the total elongation was 6270. This tubing was expanded to 0.166-in. ID and s t ra ightened. p r o c e s s resu l ted in 770 cold work , which was allowed to r e m a i n i n the material.

The u l t imate tens i le s t rength (UTS) a t room t e m p e r a t u r e , as

The expansion

The Type 316 s t a in l e s s s t ee l tubing was rece ived as fully annealed seamless tubing with a 0.166 f 0,0005-in. ID and a 0 .015 f 0.0005-in. wall th ickness . t ion was 5070.

The UTS at room t e m p e r a t u r e was 81,400 ps i , and total elonga-

The Hastel loy-X tubing was s e a m l e s s and in the annealed condition.

The UTS at r o o m t e m p e r a t u r e was 120,200 ps i , Specified dimensions of the tubing w e r e 0.166 f 0.0005-in. ID and 0.015 f 0.0005-in. wal l thickness . and the total elongation was 3370.

The Hastelloy-X-280 tubing was annealed and had been fabr ica ted Specified dimensions w e r e 0.176 2 0.001-in. by the "Weldrawn" p r o c e s s .

ID and 0.015-in. (f570) wall thickness . to ta l elongation, as de te rmined by the vendor , w e r e 121,200 p s i and 4470, respect ively.

The room- tempera tu re UTS and

The composi t ions of the four lots of tubing a r e given in Table 111. The m e a s u r e d dimensions a r e shown in Table IV .

T A B L E 111. C h e m i c a l C o m p o s i t i o n s of C l a d d i n g s

C o m p o s i t i o n (w/o )

M a t e r ial C M n Si Ni C r Mo C u C o W Fe

T y p e 3 0 4 L SS 0 . 0 3 1 .60 0 . 5 2 10 .36 18 .53 - - - - B a l . T y p e 3 16 SS 0 . 0 5 1.92 0.49 13 .17 17 .18 2 .42 0 . 2 5 - - B a l . Has t e l l o y - X 0 .13 0.50 0 . 5 3 48 .00 21 .96 8 . 6 7 - 1 . 5 7 0 . 4 3 B a l . H a s t e l l o y - X - 2 8 0 0 .10 0 . 5 8 0 . 7 4 49 .43 22 .41 8 . 6 6 - 0 . 0 7 0 . 0 1 B a l .

T A B L E IV. M e a s u r e d D i m e n s i o n s of Tub ing f o r C ladd ing

M a t e rial

Wal l O D IDa T h i c k n e s s ( i n . ) ( i n . ) ( i n . )

T y p e 3 0 4 L SS 0 .2052-0 .2056 0 .1672 0 .0190-0 .0192 T y p e 3 16 SS 0 .1953-0 .1957 0 . 1 6 6 4 0 .O 145-0 .O 147 H a s t e l l o y - X 0 .1953-0 .1969 0 .1659 0.0 147 -0 .O 155 H a s telloy - X - 2 80 0 .2065-0 .2068 0 .1764 0 .0151 -0 .0152

aAv e r a g e b y d i f f e r e n c e .

14

The tube s tock was nondestructively tes ted at ANL by eddy-cur ren t techniques. cause for re ject ion. with acetone and r insing with ethyl alcohol.

Defects with a depth of 1070 o r m o r e of the wall thickness w e r e Acceptable tubes w e r e thoroughly cleaned by scrubbing

The rod s tock used fo r the end plugs and r e s t r a i n e r s was nominally the s a m e composition as the tubing, but not usually f r o m the same heat . The rods were u l t rasonica l ly inspected for defects p r i o r to machining the end plugs and r e s t r a i n e r s .

IV. DESIGN O F THE F U E L ELEMENTS

The fuel e lements w e r e of the sealed-plenum, sodium-bonded fuel design ( s e e F ig . 3). Design da ta for each e lement a re given in Table V.

A FLAT FOR IDENT. NO.

r S E A L WELD

HELIUM

CLADDING

RESTRAINER

No LEVEL

FUEL

Na ANNULUS

SEAL WELD

Fig. 3. Schematic of Fuel Element with Restrainer

The fuel smear densi ty of the element was nominally 7570 i n all claddings except the Hastelloy-X-280, which was 6570. S m e a r densi ty i n this experiment is the ratio of the c ross -sec t iona l area of the fuel to the total c ross -sec t iona l a r e a inside the cladding. Variations i n the m e a s u r e d densi t ies of the fuel pins had only a small effect on the calculated smear densi t ies .

The gas plenum of each e lement was between 1.13 and 1.35 t i m e s as long as the fuel column (13 in . ) at room t empera tu re . The plenum volume was 153-20570 of the fuel volume at r o o m t empera tu re . Init ially, the plenum was filled with hel ium gas at sl ightly l e s s than one a tmosphere at room tempera - t u re . The plenum occupied 45-5370 of the total volume inside the cladding. height above the fuel to which the sodium was loaded was the max imum height that would not resu l t i n s t r e s s e s in the cladding g r e a t e r than the 10,000-hr c reep - rup tu re

The design

stress at 675°C. The des ign took into account fuel swelling, 2570 f iss ion- gas r e l e a s e in going to 10 a /o burnup, and t h e r m a l expansion of the components, including the sodium, which reduced the s i z e of the plenum. swelling f r o m solid f iss ion products was a s sumed to b e 2.570 p e r a tom percent burnup.

Fue l

Exper ience had indicated that the fuel would elongate i n going to the or iginal t a rge t burnup of 10 a/o. w e r e equipped with an axial r e s t r a i n e r i n the f o r m of a c ros s15 cr imped into the tubing above the leve l of the fuel column. The dis tance between

Therefore , s o m e of the e lements

15

A

the r e s t r a i n e r and the fuel column differed for different e lements . The spaces between the a r m s of the c r o s s permi t ted movement of gas o r mel ted sodium pas t the r e s t r a i n e r .

TABLE V. Des ign F a c t o r s of F u e l E lemen t s

Sodium L e v e l o v e r F u e l R e s t r a i n e r

F u e l Type of Densi ty T e m p e r a t u r e Gas -p lenum above F u e l Smear at Room Locat ion

E lemen t Cladding (70 1 (in.) Volume ( c m 3 ) ( in . )

ND28 304L SS 73 4.56 5.54 5.98 ND4 1 304L SS 73 4.53 5.56 - ND32 316 SS 74 5.28 5.24 6.90 ND43 Has te l loy-X 74 4.16 5.67 - ND2 5 304L SS 74 2.62 5.62 - ND2 7 304L SS 73 4.00 5.77 5.98 ND26 316 SS 74 5.41 5.29 - ND29 316 SS 72 5.41 5.30 - ND30 316 SS 72 5.41 5.30 - ND3 1 316 SS 74 5.30 5.23 6.90 ND33 H a s te l loy-X 73 4.09 5.66 - ND34 Has te l loy-X 73 4.25 5.70 ND3 5 Has te l loy-X 75 4.08 5.63 5.48 ND37 Has te l loy- X- 280 66 2.66 7.02

ND44 Hastel loy-X-280 67 2.66 7.03 - ND39 Hastel loy-X- 280 66 2.67 6.99 3.79

Note: Sodium leve ls d e s i r e d w e r e : Type 304 SS, 4.4 ? 0.1 in . ; Type 316 SS, 5.1 k 0.1 i n . ; Has te l loy-X, 3 .9 ? 0.1 in . ; and Hastel loy-X-280, 2.4 + 0.1 in. w e r e m a d e i n ND25 and 27. O t h e r d i f fe rences w e r e technica l v a r i a t i o n s . L e v e l s d e t e r m i n e d by radiography.

E r r o r s

E i ther of two procedures was used to a s semble the fuel e lements i n a glovebox, depending on whether o r not the element was to be equipped with an axial r e s t r a i n e r . In each e lement without a r e s t r a i n e r the bottom end plug was welded to the tubing and the weld was leak- tes ted with a hel ium m a s s spec t romete r . Acceptance was based on no hel ium detection on the m o s t sens i t ive de tec tor s c a l e , which had the capabili ty of detecting a s tan- da rd leak of 2.8 x cm3/sec . A prede termined amount of extruded s o - dium was in se r t ed into the cladding and the fuel pin was placed on top of the sodium. by weight. The par t ia l a s sembly was placed i n a ve r t i ca l posit ion, heated to 150°C to m e l t the sodium, and then v ibra ted to se t t l e the pin to the bot- t om of the cladding. Because of the negative glovebox a tmosphere , the hel ium in the plenum was at sl ightly l e s s than a tmospher ic p r e s s u r e .

Analysis of the sodium showed an oxygen content of 30-50 ppm

The top end plug was then welded to the cladding.

Each fuel e lement that was to be equipped with a r e s t r a i n e r was a s sembled by cr imping the r e s t r a i n e r into position , inside of the cladding,

then loading the fuel pin and p a r t of the sodium through the bottom end. The bottom end plug was welded shut and the remainder of the sodium loaded through the top. 150"C, af ter which hel ium was applied under p r e s s u r e to fo rce the sodium above the r e s t r a i n e r into the lower p a r t of the cladding, The top end plug was then welded shut and the completely sea led element checked with a hel ium leak de tec tor .

The tube was evacuated to 200 p m and heated to

Any element that leaked was r epa i r ed .

The fuel e lements w e r e subjected to one o r m o r e bonding cycles a t The bottom of each element was rapped 1000-2000 t i m e s by an air- 500°C.

activated piston to remove bubbles f r o m the sodium and to promote bonding. The annular sodium bond was inspected at 150°C by eddy-cur ren t techniques. Rebonding was insti tuted i f voids 1/16 in. o r g r e a t e r i n d i ame te r w e r e detected along the fuel column.

The overa l l length, outside d i ame te r , weight, and volume of each e lement w e r e m e a s u r e d . These data a r e tabulated i n Table VI.

T A B L E VI. P r e i r r a d i a t i o n M e a s u r e m e n t s of F u e l E l e m e n t s

Type of Overa l l F u e l Cladding Length ODa Weight Volume

E l e m e n t M a t e r i a l ( in . ) ( in . ) ( P ) ( cm3)

ND28 ND4 1 ND32 ND43 ND25 ND27 ND26 ND29 ND30 ND3 1 ND3 3 ND34 ND3 5 ND37 ND39 ND44

304L SS 304L SS 316 SS

H a s te l loy- X 304L SS 304L SS 316 SS 316 SS 316 SS 316 SS

H a s telloy-X H a s te l loy-X H a s telloy-X

Hastel loy-X-280 Has te l loy- X- 2 80 Hastel loy-X-280

33.938 33.938 33.931 33.930 33.938 33.938 33.928 33.939 33.935 33.933 33.939 33.933 33.931 33.934 33.932 33.929

0.2053 0.2054 0.1953 0.1953 0.2056 0.2052 0.1954 0.1957 0.1955 0.1955 0.1963 0.1959 0.1969 0.2065 0.2068 0.2065

106.668 106.321 94.878 95.991

104.108 104.971 93.252 91.708 91.884 93.879 94.442 93.884 96.604 97.653 98.082 97.691

18.179 18.193 16.511 16.484 18.198 18.185 16.512 16.512 16.522 16.531 16.661 16.601 16.652 18.432 18.440 18.431

aAverage of 24 m e a s u r e m e n t s of e a c h e lement . Average f o r Type 304L SS,

0.2054 :!:E!::; fo r Type 316 SS, 0.1955 ?::!toof; fo r Has te l loy-X,

0.1961 ?::E::;; and f o r Has te l loy-X- 280, 0.2066 ?::E;:! in.

V. IRRADIATION IN EBR-I1

Each of the 16 fuel e lements was encapsulated in a s tandard Type 304 s ta in less s t ee l capsule for i r rad ia t ion in a Type 19A EBR-I1 exper imenta l subassembly (XAO7). A V-2OTi cup se rved a s a b a r r i e r between the fuel e lement and the capsule i n the event of e lement fa i lure . Sodium was the hea t - t r ans fe r medium i n the capsule. The sodium level was 8.4-10.4 in.

17

above the top of the fuel i n the e l emen t , but below the top of the element . The capsule gas plenum, filled with he l ium, was s ized such that the calcu- la ted plenum p r e s s u r e was 47-48 ps i during ir:radiation at 525°C (ave rage ) . The capsules w e r e bonded by heating a t 500°C and lightly tapped to se t t le the components. and fo r the posit ions of the fuel e lement and the V-20Ti cup. inspect ion was also used for void detection. u sed to check the he rme t i c i ty of the capsule welds. su l e s w e r e reworked and reinspected.

Each capsule was radiographical ly inspected for voids

A he l ium leak de tec tor was Unsat isfactory cap-

Eddy-cu r ren t

The XA07 exper imenta l subassembly was i r r ad ia t ed in posit ion 3 ,

The operat ing Numerous power

row 4 , s ec to r D of EBR-I1 f r o m October 27, 1965 to December 1 2 , 1966, f o r a total exposure of 7950 MWd (41.570 operat ing efficiency). power level was 45 MW for 90-9570 of the total exposure. changes o c c u r r e d during the i r r ad ia t ion of this subassembly . power d e c r e a s e s at the no rma l shutdown r a t e o f -0.68 MW/min o c c u r r e d ; of t hese 15 w e r e f r o m 45 to 0 MW, and 19 involved power changes of 20 MW o r g r e a t e r . T h e r e w e r e 11 fast shutdowns at a. r a t e of 10 MW/min; six w e r e f r o m full power (45 MW), and nine involved power d e c r e a s e s of 20 MW o r g r e a t e r . P o w e r changes a t a s c r a m r a t e of -45 MW/sec number 58; 36 w e r e f r o m 45 to 0 MW, and 41 involved power changes of 20 MW o r g r e a t e r . total number of shutdowns f r o m full power was 57; the total number of power d e c r e a s e s of 20 M W o r g r e a t e r was 69. c r e a s e at s t a r t u p was -0.68 MW/min.

Forty- two

The

The max imum r a t e of power in-

The in le t t e m p e r a t u r e of the sodium to th.e r e a c t o r c o r e was 700°F (371"C), and the flow r a t e of sodium through the subassembly was 16.2 gpm, which was obtained by s iz ing the or i f ice a t the inlet of the subassembly . This r a t e was based on a m a x i m u m cladding t e m p e r a t u r e of 625°C.

Sixteen of the 19 posit ions i n the subassembly were occupied by encapsulated fuel e lements ; the o ther t h r e e locations contained capsules with s t r u c t u r a l m a t e r i a l s . the r e a c t o r , where the cen te r of the subassembly was located 15.59 c m f r o m the r e a c t o r cen te r . The r ad ia l d i s tances f r o m the center of the r e - ac to r to the n e a r e s t and f a r thes t posit ions in the subassembly w e r e 13.13 and 18.07 c m , respect ively. f a r t h e s t posit ions a r e given i n Table VII.

The subassembly was placed in posit ion 4D3 i n

The f iss ion r a t e s for the n e a r e s t , c e n t e r , and

TABLE VII. Fission Rates in EBR-I1 Core Location 4D3 at 45 MW

Fission Rates for Different Subassembly Locations with Respect to Reactor Center (f iss ions/g-sec)

Nearest , Center, Farthest , Mat e r i a1 13.13 cma 15.59 cma 18.07 cma

Fuel

2 3 5 ~ 0.796 x 1013 0.744 1013 0.698 x lOI3 2 3 8 ~ 0.522 x 10'' 0.483 x 10" 0.424 x 10" Pu 0 .962 x 1013 0.900 x 1013 0.845 1013

aDistance from reactor center.

18

M a x i m u m L i n e a r H e a t M a x i m u m

F u e l G e n e r a t i o n Spec i f ic P o w e r E l e m e n t (kW/ f t 1 (kW/kg U t P u )

The f i ss ion r a t e s , exposure t ime , and fuel compositions w e r e used

The max imum values

T h e s e calculated burnup values

to calculate burnup of the heavy e lements in a tom percent . Table VI11 gives the r e su l t s fo r the different subassembly posit ions. of burnup in terms of MWd/T of heavy a toms ranged f r o m 34,000 to 38,000, the average va lues f r o m 30,000 to 34,000. are higher than those obtained f r o m pos t i r rad ia t ion ana lyses . 4

M a x i m u m L i n e a r H e a t Max i m u m

F u e l G e n e r a t i o n Spec i f ic P o w e r E l e m e n t (kW/ f t ) (kW/kg U t P u )

T A B L E VIII. S u b a s s e m b l y P o s i t i o n s a n d C a l c u l a t e d B u r n u p

ND2 8 1 0 . 5 233 ND41 1 0 . 3 228 ND32 10.0 22 1 ND43 10 .2 226 ND2 5 9.2 218 ND27 9 . 4 222 ND26 9.1 213 ND29 9 . 3 225

Burnupb M a x T o t a l ( 4 0 ) D i s t a n c e f r o m F u e l C o r e C e n t e r M a x F i s s i o n R a t e a F i s s i o n s

ND30 10 .1 244 ND3 1 9 . 8 230 ND33 9.6 232 ND34 9 .8 237 ND3 5 10 .1 236 ND37 10.1 241 ND39 10.0 237 ND44 9.7 229

E l e m e n t ( c m ) (10” f i s s / c m 3 - s e c ) (10“ f i s s / c m 3 ) M a x Avg

ND28 ND41 ND32 ND43 ND2 5 ND27 ND26 ND29 ND30 ND3 1 ND33 ND34 ND3 5 ND37 ND39 ND44

14.80 15.59 16.83 16 .05 17.30 16.59 18 .07 16 .14 13.13 15.22 14.96 14.02 14.36 13.55 14.07 15 .33

1 .104 1.081 1 .048 1.070 0.967 0.985 0.952 0.972 1.056 1.028 1.004 1.029 1.056 1 .059 1.047 1.016

16.1 15.7 15.3 15 .6 14.1 1 4 . 3 13.9 14 .2 1 5 . 4 15.0 14.6 15.0 1 5 . 4 1 5 . 4 15.2 14 .8

4 .5 3 .9 4 . 4 3 . 8 4.2 3.7 4.3 3 . 8 4.2 3.7 4.3 3.7 4 .1 3.6 4.3 3 . 8 4 .7 4.1 4 . 4 3.9 4.5 3 .9 4.6 4 .0 4.5 4.0 4.6 4.0 4 .6 4 .0 4 . 4 3.9

aAt r e a c t o r m i d p l a n e . bOf h e a v y a t o m s .

.. . . .

19

A

t

f r o m 213 to 244 kW/kg (U t P u ) and the m a x i m u m l inea r hea t -genera t ion rates f r o m 9.1 to 10,5 kW/ft. The max imum volumetr ic hea t -genera t ion r a t e s w e r e 675-783 ca l / cm3-sec .

The max imum values of the hea t flux at the inner su r face of the

The l a r g e r ins ide d i a m e t e r of the Hastelloy-X-280 cladding r e - claddings with a nominal 0.166-in. ID ranged f r o m 710,000 to 820,000 Btu/ f t2 -hr . sul ted in a m a x i m u m hea t flux of about 510,000 Btu/f t2-hr .

The fuel and cladding t e m p e r a t u r e s w e r e calculated f o r the condi- The ini t ia l min i - t ions exis t ing at the beginning of and during i r rad ia t ion .

m u m , max imum, and ave rage t e m p e r a t u r e s of the fuel and the cladding for each e lement a r e given i n Table X. fo r ND35, a typical e lement , a r e shown in Fig. 4. The m a x i m u m t e m p e r a - t u r e o c c u r s in the center of the fuel at about 70% of t he height of the column. The m a x i m u m t e m p e r a t u r e of the inside d i a m e t e r of the cladding o c c u r s a t a point corresponding to -90% of the height of the fuel column,

The axial t e m p e r a t u r e dis t r ibut ions

The fuel t e m p e r a t u r e s during i r r ad ia t ion w e r e calculated by con- s ide r ing the effect of fuel porosi ty , developed during i r r ad ia t ion , on the t h e r m a l conductivity of the fuel. lat ing t h e r m a l conductivity and porosi ty in oxide fuel provided a r easonab le

An expres s ion used by Bel le i6 f o r r e -

TABLE X. Init ial Fue l and Cladding 'Tempera tu res

C 1 adding Fue l T e m p e r a t u r e s ("C) T e m p e r a t u r e s ("C )

Element Min Max Volume Avg Min Max Avg Fue l

ND28 ND41 ND32 ND43 ND25 ND27 ND26 ND29 ND30 ND3 1 ND33 ND34 ND35 ND37 ND39 ND44

49 1 762 48 9 753 47 8 731 48 2 7 42 476 713 47 8 720 46 8 697 47 0 705 47 8 733 476 724 47 5 7 19 47 8 728 48 1 736 479 735 47 7 731 47 4 719

6 40 633 617 623 599 665 585 59 2 616 608 609 612 618 619 614 604

44 1 652 572 440 647 572 43 8 629 557 439 638 564 43 3 618 547 43 4 622 5 50 43 2 60 5 53 5 43 3 61 1 541 43 8 631 558 43 7 624 552 43 5 621 551 437 ' 627 555 43 8 634 561 43 9 632 559 43 9 629 557 43 6 622 550

FUEL HEIGHT, cm

AT HEIGHT OF 23 an I ON FUEL COLUMN I

Fig. 4. Axial Temperature Distribution i n Fuel Element ND35

corre la t ion with experimental resu l t s on i r r ad ia t ed U-5Fs fuel.17 is

The equation

1 - P 1 -t PPI

kp = ko ~

where ko is the the rma l conductivity without porosi ty , p is the fract ion of porosi ty , and /3 = 0.5 is a fac tor that t akes into account the a s sumed spher i - cal shape of the voids. Because of the observed complete swelling of the fuel into the sodium annulus, the U-15Pu-1OZr alloy was a s sumed to have 2570 porosity. F o r the t empera tu re calculations, the porosi ty was a s sumed to be uniformly di s t r ibut ed . distributions in both the axial and the rad ia l direct ions are shown in Figs . 4 and 5, respect ively. After the fuel has

The c a1 cul at e d t emp e r a tu r e

swelled to the cladding, the fuel su r f ace t empera tu re is a s sumed to be that of the inner cladding su r face . These calculated t empera tu res are based on the f i ss ion rates in EBR-I1 and on the a s sumed coolant flow rates through the subassembly.

The calculated max imum in t e r - nal p r e s s u r e s in the e lements , the cladding hoop s t r e s s e s , and the thermal s t r e s s e s at the end of i r rad ia t ion are given in Table XI. The p r e s s u r e s range f r o m 410 to 575 ps i , with cor respond- ing hoop s t r e s s e s of 2320 and 3300 psi . The in te rna l p r e s s u r e s w e r e calculated, based on solid f iss ion-product swelling of 2.570 pe r a tom percent burnup and on 100% fiss ion-gas r e l ease . To calculate the max imum possible s t r e s s on the cladding, total gas r e l e a s e was assumed. The actual g a s - r e l e a s e values w e r e of

I I I I I I I

600 0 IO 20 30 40 50 60 70 80

RADIAL DISTANCE FROM FUEL CENTER, mils

Fig. 5. Radial Temperature Distribution in Fuel Element ND35

the o r d e r of 63%. Averages of the axial t empera tu res during i r rad ia t ion fo r the fuel, the sodium, and the cladding w e r e used in the calculations.

.

The claddings w e r e exposed to fast fluences ranging f r o m 2.9 x 10'' to 3 .4 x lo'' nvt at the r e a c t o r midplane (maximum). calculated f r o m fluxes measu red by personnel at EBR-11.

These fluences w e r e

M a x E l e m e n t Cladding Cladding P r e s s u r e Hoop T h e r m a l

F u e l at T e m p S t r e s s S t r e s s E l e m e n t (Ps i ) (Ps i ) (Ps i )

ND2 8 ND4 1 ND32 ND43 ND2 5 ND2 7 ND2 6 ND29

M a x E l e m e n t Cladding Cladding Pres s u r e Hoop T h e r m a l

F u e l a t T e m p S t r e s s S t r e s s E l e m e n t (ps i ) (Ps i ) (Ps i )

562 550 572 53 5 49 1 482 519 529

2 540 2470 3300 30 10 2170 2150 2960 2990

14,500 14,100 12,600 11,500 12,700 12 ,900 11,500 11,700

ND30 ND3 1 ND33 ND34 ND3 5 ND3 7 ND39 ND44

57 5 564 506 517 534 42 8 42 5 410

3300 3230 2750 2850 2860 2 440 2400 2320

12 ,700 12 ,400 10,800 11,100 11,400 10 ,400 10,300 10,000

aIn i t ia l p r e s s u r e s at t e m p e r a t u r e : c a p s u l e , 47-48 p s i , and e l e m e n t , 45-46 ps i .

VI . POSTIRRADIATION EXAMINATION

A. Disassemblv of the Capsules

Four teen o f the 16 capsules w e r e visual ly examined in the hot cell . Capsules ND31 and 33 w e r e not examined. bowed, one as much as 3 / 4 in . , although the average was -1/8 in. successfu l a t tempt was made to r e l a t e the d i rec t ion of bowing to the loca- t ion of the capsule in the subassembly . f r o m remote mechanical handling of the long (40 in . ) i r r ad ia t ed capsules .

Many of the capsules w e r e An un-

The bowing may have resu l ted

The 14 capsules w e r e punctured i n the plenum region to collect the No radioactivity was found in any of the gas s a m p l e s , which indi- gases .

cated none of the elements had failed.

The top end of each capsule was seve red to provide a c c e s s to the elements . The sodium in the capsules was then mel ted , and the fuel e le - ments removed and then cleaned with butyl alcohol to remove any clinging sodium.

B. Dimensions of the Fuel Elements

Visually, the 14 fuel e lements appeared in excellent condition, with no evidence of dis tor t ion o r damage to the cladding. examination did revea l sl ight changes in the geometry .

However, detailed

P r i o r to i r rad ia t ion the e lements had been mounted in a la the and bowing was m e a s u r e d with a dial indicator . bowing, 0 .008-0 .070 in , (average 0.02 in . ) . of bowing w e r e made with a r u l e r graduated in 1/64-in. i nc remen t s . max imum height of the a r c h made by the bow was m e a s u r e d , and an in- c r e a s e in bowing was observed. The maximum bowing after i r rad ia t ion

All the e lements showed s o m e Pos t i r rad ia t ion measu remen t s

The

22

was 0.28 in . , the a v e r a g e being 0.13 in , re la ted with the posit ion of the e lement i n the subassembly .

These r e su l t s could not be c o r -

The length of each fuel e lement was m e a s u r e d in a f ixture i n which The s m a l l e s t division a s tandard length was used to z e r o a d ia l indicator .

on the dial was 0.01 mm. m i l l i m e t e r s to inches , and are compared with p re i r r ad ia t ion lengths i n Table XII.

The pos t i r rad ia t ion lengths w e r e converted f r o m

TABLE XII. Dimensional Changes of the I r r ad ia t ed Fue l Elements

Average Diametra l Changes Length

All Along Fue l Above Fue l Changes d Fue l Values a Column b ColumnC

El e m e nt (mils ) (mils) (mils) mils 70

ND28 ND41 ND32 ND 43 ND25 ND27 ND26 ND29 ND30 ND3 1 ND33 ND34 ND35 ND37 ND39 ND44

Av e r ag e

t o . 1 +o. 1 t 0 . 3

0 .0 -0.2 t o . 1 -0 .1

0 , o t 0 . 2

- -

t 0 . 2 -0 .8

0 .0 -0.1 t 0 . 2

t 0 . 5 $0.4 t O . 6 t 0 . 2 -0.2 t 0 . 2

0 , o 0 .o

t 0 , 3 - -

t0.3 -0.5 t 0 .2 t 0 . 2 t 0 . 5

0 .o t 0 . 2

0 .0 0 .0

$0.1 -0.2 -0.3 $0. 1 - 0 . 2

0 .o 0 . 0 - -

t o . 1 - 1 .o -0 .2 -0,2 -0.1

- 0 , l -

t 1 6 $1 8 t10 t 3 8 t 1 5 t 1 5 t12 t 1 2 $13 - I

342 t10 t 1 0

$8 +44

t 1 9 -

0.05 0 .05 0.03 0.11 0 .04 0.04 0 . 0 4 0.04 0 . 0 4

- -

0.12 0.03 0.03 0 .02 0.13

0 , 0 6 -

"Twenty-four d i a m e t r a l m e a s u r e m e n t s along fuel e lement at

bTen d iame t ra l m e a s u r e m e n t s along fuel column, CFourteen d i a m e t r a l m e a s u r e m e n t s along e lement above fuel

3-in. in te rva ls and two rotat ions, 0" and 90".

column. All m e a s u r e m e n t s are compared with or iginal d iame- tral m e a s u r e m e n t s ( s e e Table VI).

dBased on or iginal length m e a s u r e m e n t s ( s e e Table VI ) .

The d i a m e t e r s of the i r r a d i a t e d fuel e lements w e r e m e a s u r e d with a 1-in. m i c r o m e t e r graduated in inc remen t s of 0.0002 in. w e r e es t imated to 0.0001 in .

The readings Two readings , 90" apa r t , w e r e tak.en about

eve ry 3 in. along the length, s t a r t i ng 1 /2 in, above the bottom plug. The differences between the p r e - and pos t i r rad ia t ion ave rage d i a m e t e r s a r e given in Table XI1 for measu remen t s (a) along the en t i re length of the ele- men t , (b) ove r the length of the fuel column, and (c) along the length above the fuel column. points ranged f r o m t O . l to t1.0 mils (average t0 .6) . The g r e a t e s t d iame- tral differences w e r e , in genera l , along the upper half of the fuel column.

The max imum d iame t ra l differences at corresponding

C . Weights and Volumes of t he Fuel Elements

The fuel e lements w e r e weighed t o the n e a r e s t m i l l i g r a m in the ni t rogen a tmosphere of the hot cell. l o s s of f r o m 9 to 16 m g , presumably because of sodium corros ion . weight l o s s e s p e r unit su r f ace a r e a of the e lements a r e given i n Table XI11 The volumes of the fuel e lements w e r e m e a s u r e d by i m m e r s i o n in carbon te t rachlor ide . Table XI11 shows the pos t i r rad ia t ion volumes and the volume changes, along with the weight changes.

Each element experienced a weight The

Volume i n c r e a s e s ranging f r o m 0.1 to 0,370 w e r e found.

T A B L E XIII. Weight and V o l u m e C h a n g e s of I r r a d i a t e d F u e l E l e m e n t s

F u e l E l e m e n t

P o s t i r r a d i a t i o n P o s t i r r a d i a t i o n AVa V o l u m e a Weigh t V o l u m e I n c r e a s e I n c r e a s e

Weight Loss

(9 ) m g a m g / d m z b ( c m 3 ) (cm3) (70)

ND28 ND4 1 ND32 ND43 ND2 5 ND2 7 ND26 ND29 ND30 ND3 1 ND33 ND3 4 ND3 5 ND3 7 ND29 ND44

106.652 106.309

94 .868 95 .979

104.096 104.958 93 .243 9 1.698 9 1.870

- -

93.871 96 .592 97.641

97 .682 98 .070

16 12 10 12 12 13

9 10 14 - - 1 3 12 12 12 9

1 1 . 5 8 .7 7 .6 9 .1 8 .6 9 . 4 6.8 7 .6

10.6 -

9.8 9 .0 8.6 8 .6 6 . 5

18 .224 18.237 16 .538 16.499 18 .232 18.240 16 .531 16.539 16.543

16 .621 16 .684 18 .472

18 .465 18.476

0 . 0 4 5 0 . 0 4 4 0 .027 0.0 15 0 .034 0 .055 0.019 0 .027 0 .021

- 0.020 0 .032 0.040 0 .036 0 . 0 3 4

0 . 2 5 0 . 2 4 0 . 1 6 0 .09 0 .19 0.30 0 .12 0 .16 0 . 1 3

- 0 . 1 1 0 .19 0 . 2 2 0.20 0 .18

-

a c o r n p a r e d wi th p r e i r r a d i a t i o n data i n T a b l e VI. b L o s s p e r s q u a r e d e c i m e t e r of f u e l - e l e m e n t s u r f a c e .

D. Dimensions of the Fuel P ins

The 16 i r r ad ia t ed e lements w e r e examined by neutron radiography while they w e r e still encapsulated. The neutron radiographs provided information on fuel growth, and on separa t ions and displacements i n the fuel column. A length s tandard with m a r k e r s at intervals of 3 + 1/64 in. was exposed along with the capsules . A typical neutron radiograph is shown in Fig. 6 along with a gamma scan.

23

Fig. 6 . Gamma Scan and Neutron Radiograph of Fuel Column ND25. (a) Gamma scan; (b) Neutron radiograph.

2 5

The fuel columns had moved off the bottom plug 0.06-0.37 in . , with an average of 0.20 in. ( s e e Table XIV). separa t ions in the fuel columns of each of 14 of the 16 e lements . two fuel columns had no separa t ions . 22 . Somet imes toward the center of the column the re was necking between the par ted sect ions ( s e e Fig. 6 ) . with r e spec t to position along the fuel column. a lent below the midlength of t he fuel column.

The re w e r e f r o m one to t h r e e axial The other

The total number of separa t ions was

- They var ied f r o m ba re ly vis ible c racks to a gap of a lmost 1/16 in.

F igu re 7 shows the dis t r ibut ion of separa t ions Separat ions w e r e m o r e prev-

T A B L E XIV. L e n g t h C h a n g e s of I r r a d i a t e d F u e l C o l u m n s ~

S u m of A c t u a l E f f e c t i v e

F u e l of B o t t o m S e p a r a t i o n s L e n g t h a F u e l b I n c r e a s e ' D i s t a n c e Off F u e l C o l u m n F u e l A L of c u e 1 C o l u m n

E l e m e n t P l u g ( i n . ) ( in . ) ( i n . ) (70) (70)

ND28 0 . 1 2 0 . 0 2 13 .01 t o . 1 1 . 2 ND4 1 0 . 1 9 0.0 1 13 .10 t 0 . 8 2 . 3 ND32 0 . 1 9 0.04 1 3 . 1 1 t 0 . 9 2 .6 ND43 0 . 3 6 0.00 12.91 -0.7 2.1 ND2 5 0 . 1 2 0.06 13 .40 t 3 . 1 4 . 5 ND27 0 .23 0 . 0 2 13.37 t 2 . 9 4 . 8 ND26 0 . 1 3 0 . 0 3 13 .79 t 6 . 1 7 . 3 ND29 0 . 1 3 0 . 0 3 13 .71 t 5 . 5 6 .7 ND30 0 . 2 5 0 . 0 2 13 .13 t 1 . 0 3 . 1 ND3 1 0 . 1 9 0.05 1 3 . 1 8 t 1 . 4 3 . 2 ND33 0 . 0 6 0 . 0 2 13 .17 t 1 . 3 1 . 9 ND34 0 . 3 2 0 . 0 1 13 .02 t 0 . 2 2.7 ND3 5 0 . 3 7 0.00 12.87 - 1 .o 1 . 8 ND37 0 . 1 3 0 . 0 2 1 3 . 2 5 t 1 . 9 3 . 1 ND39 0 . 0 8 0.02 13 .25 t l . 9 2.7 ND44 0 . 2 5 0 . 0 3 13 .16 $1 .2 3 . 4

a L e s s s e p a r a t i o n s . b B a s e d o n o r i g i n a l l e n g t h of f u e l (see T a b l e 11). C P e r c e n t a g e Effective I n c r e a s e = -

100 x ( F u e l L e n g t h t D i s t a n c e Off P l u g t S e p a r a t i o n s - O r i g i n a l L e n g t h ) O r i g i n a l L e n g t h

IO I

XI 22 SEPARATIONS 9- &I IN 14 OF 16 G: ELEMENTS 8 -

FRACTIONAL DISTANCE FROM BOTTOM OF FUEL COLUMN

Fig. 7. Axial Distribution of Separations in Fuel Columns

Changes in the lengths of the fuel columns had a l so occur red , as shown in Table XIV. The two fuel columns that had no separa t ions dec reased in length. These w e r e ND43 and 35, one of each fuel alloy composition, but both clad inHastel loy-X. All the o ther fuel columns inc reased in length, even after allowance for the b reaks . Because of the po- tent ia l associat ion with react ivi ty effects in the r e a c t o r , the percentage effective i n c r e a s e of the fuel columns has a lso been given in Table XIV. This fac tor includes the lift-off f r o m the bot tom plug, the total fuel-column separa t ion , and the actual length change of the fuel column. tor t ion was noted in s o m e of the fuel columns, and was subsequently confirmed by metal lography.

End d i s -

26

I n the neut ron rad iographs the fuel appeared to have swel led to the inside of the cladding, even fo r the l a r g e r ins ide d i a m e t e r (0.176 in.) of the Hastelloy-X-280 tubing (ND37, 39, and 44). However , the resolut ion of the rad iographs was inadequate f o r the detect ion of small separa t ions between fuel and cladding.

The fuel vo lumes w e r e calculated f r o m the lengths m e a s u r e d on the These vol- neutron rad iographs and the ins ide d i a m e t e r s of t he claddings.

u m e s and the calculated volume i n c r e a s e s a r e shown i n Table XV. The volume i n c r e a s e s ranged f r o m 32 to 46% i n e lements that had or iginal smear dens i t ies of f r o m 72 to 757'0 (0.166 in. ID) and f r o m 52 to 567'0 i n e lements with s m e a r dens i t ies of f r o m 66 to 677'0 (0.176 in. ID).

TABLE XV. Calculated Volume I n c r e a s e s of I r r ad ia t ed Fue l s

Volume of I r r a d i a t e d Volume

Fue l Fuel" I n c r e a s e AV SFPC CV FGd Element ( cm3 ) ( 7 0 ) (70 1 ( 7 0 1

ND28 ND41 ND32 ND43 ND25 ND27 ND26 ND29 ND30 ND3 1 ND33 ND34 ND3 5 ND37 ND39 ND44

4.683 4.714 4.671 4.574 4.823 4.812 4.915 4.885 4.679 4.697 4.666 4.613 4.560 5.308 5.308 5.272

37 38 37 34 40 40 43 46 40 37 40 38 32 55 56 52

10 10

9 10

9 9 9

10 10 10 10 10 10 10 10 10

27 28 28 24 31 31 34 36 30 27 30 28 2 2 45 46 42

'Calculated f r o m length of i r r a d i a t e d fuel and inside d iam-

bBased on or ig ina l volume ( s e e Table 11). 'SFP, sol id f i s s ion products--2.570 vol i n c r e a s e p e r a / o

dFG, " f i ss ion gas

eter of cladding.

burnup.

differ ence including mechanica l tear ing- -by

A

E. Axial Gamma Pro f i l e

The fuel e lements w e r e scanned f o r dis t r ibut ion of total g a m m a radiat ion between 0.1 and 2.0 MeV along the length of the fuel column, typical g a m m a s c a n t r a c e , along with a neutron radiograph, is shown in Fig. 6. The g a m m a scans tend to skew slightly toward the bottom of the fuel column. The skew was least for e lements on the s ide of the subas- s embly c loses t to the r eac to r center . The l a r g e separa t ions in the fuel column are evident in the t r a c e s , but the s m a l l e r separa t ions s e e n on the neutron radiographs a r e not always apparent . Some separa t ion- l ike indications appear on the g a m m a t r a c e without corresponding evidence on the radiographs.

A

F. Determinat ion of Sodium Level

The condition of the sodium and its height above the top of the fuel w e r e de te rmined by eddy- cur ren t techniques. bubbles and separa t ions in the sodium column above the fuel i n s e v e r a l e lements . the e lements w e r e vibrated 15-30 m i n in an at tempt to consolidate the sodium. m e n t did not improve two of the e lements . mined by the eddy-cur ren t tests indicated that the re was no loss of sodium f r o m the e lements . Although i r r e g u l a r and d i s s imi l a r t r a c e s w e r e often obtained along the fuel columns, the r e su l t s could not be co r re l a t ed with separa t ions in the fuel. dicated that a sodium annulus might still be p re sen t and that the fuel probably had not swelled to the cladding.

Init ial examination revea led

The top portions of the 14 e lements w e r e heated to 150°C, and

Improvement was noted in s o m e c a s e s , but even a second t r e a t - The sodium levels as d e t e r -

The t r a c e s near the bottom of a few e lements in-

G. Analvsis of F i s s ion Gas

Eight fuel e lements (ND28, 32, 43 , 25, 26 , 30, 35, and 37) w e r e punctured in the plenum at room t e m p e r a t u r e to col lect g a s e s . Neutron radiography had indicated that four of t hese e lements (ND28, 32, 35, and 37) each had a plug of sodium in the plenum. After exposure to the hot- cell a tmosphe re (ni t rogen) , the plenum sect ion of each of the four punc- tu red e lements was heated to -150°C to melt the sodium. The additional gases r e l eased w e r e collected sepa ra t e ly and analyzed with a mass spec- t rograph . or iginal f i l l -gas in the plenum). the e lements contained about 1070 of the theore t ica l amount of krypton and xenon produced, and about 1-270 of the amount of hel ium, The total quanti- t i e s of the three g a s e s a r e given i n Table XVI. The f iss ion-gas r e l e a s e is compared with o ther data" in Fig. 8.

The gases of i n t e re s t w e r e xenon, krypton, and he l ium (the The additional gas collected on heating

T A B L E XVI. F i s s i o n G a s e s (Xe t K r ) and H e l i u m Col l ec t ed from P l e n u m s of I r r a d i a t e d F u e l E l e m e n t s

H e l i u m F i s s i o n G a s e s

R e c o v e r e d In i t ia l R e c o v e r e d C a l c Con ten tC F u e l E l e m e n t ( c m 3 a t STP) c m 3 a t STP T O (om3 a t S T P ) c m 3 a t STP T O

26.7 54 - - - 3.3 7 - -

30.0 61 5.1

3 . 8 74 2 0.1

3.9 76 - -

T o t a l 49.1

ND32a - ND32b

T o t a l 46.6

- - 25.6 55 -

12

31.0 67 4.8

- - - - 5.4

1 . 8 38 0.1 2

1 .9 40 - -

ND43 47.6 ND25 43.0 ND26 42.3 ND30 47.0

29.6 62 5.2 25.8 60 5.2 26.5 63 4.9 '

28.3 60 4.9

4 .5 87 5.0 96 3.8 78 4.2 86

ND35a - ND35b -

T o t a l 47.0 -

27.4 58 5

29.6 63 5.2 - - 2.2 -

4.2 81 0 . 1 2

4 . 3 83 - -

ND37a - ND37b

To ta l 47.1 -

30.5 65 - 3.0 6 -

33.5 71 6 . 4 - - -

5.7 89 0 . 1 2

5 .8 91 - -

a 0 r i g i n a l s a m p l e . bAdditional s a m p l e ob ta ined by hea t ing g a s p l e n u m t o m e l t s o d i u m plug. C B a s e d o n p l e n u m v o l u m e and glovebox p r e s s u r e and t e m p e r a t u r e .

H. Metal lographic Examinat ion of the F u e l gOr----l T r a n s v e r s e sec t ions w e r e made at

var ious locations along the lengths of three e lements of U - 15Pu-9Zr clad respec t ive ly with Type 304 s t a in l e s s s t ee l , Type 316 s ta in- less s t ee l , and Hastelloy-X (ND28, 32, and 43), and two e lements of U-14Pu-12Zr clad r e s p e c - t ively with Type 316 s t a in l e s s s t ee l and Hastel loy-X (ND30 and 35). longitudinal metal lographic samples w e r e p r e - pared . nated with epoxy to r e t a in the fuel during the polishing operat ion. 1 - p m diamond pas te .

T r a n s v e r s e and

The mounted spec imens w e r e impreg-

The final polish was with The as-pol ished s u r -

BO

70

'. 60 W In

W J so W K

a

340 z In E 30 0

V U-Fa (Ret. I81

0 U-Pu-FS (Ret. 181

0 U-Pu-Zr (EBR-111 PO

0 U-Pu-Zr (CP-SI 10

0 PO 40 60 80 100 120

FUEL VOLUME INCREASE, % faces w e r e examined at magnifications of 15X and f r o m 250X to 800X.

Fig. 8. Effect of Swelling in Metallic Fuels on Fission-gas Release

A series of sec t ions f r o m ND35 (U-14Pu- 1 2 Z r , Hastelloy-X cladding at 15Xmagnification

is shown in Fig. 9 as typical of the s t r u c t u r e of the fuel e lements examined, Two zones are p resen t in the cooler sect ions of the e lement (nea r the bot tom), whereas i n the hot te r sec t ions , t h r e e zones exis t . zone formation is cha rac t e r i s t i c of both fuel compositions.

t ions a c r o s s the fuel column a r e shown in Fig. 11. m i c r o s t r u c t u r e s of the fuel are shown in F igs . 1 2 - 1 7 . ample of an etched spec imen showing the banded appearance of a white phase at the boundary between the middle and center zones. the etching cha rac t e r i s t i c s of this phase a r e similar to those of t he c (U,Pu) phase in uni r rad ia ted U - P u - Z r .

c

This pa t te rn of The top and

. bottom ends of the fuel a r e d is tor ted , as shown in Fig. 10. Typical s e p a r a - Both typical and atypical

F igu re 16 is an ex-

The morphology and

Beta- g a m m a and alpha autoradiographs w e r e made f r o m the mounted and polished spec imens Typical be ta -gamma autoradiographs for sect ions along the length of ND35 a r e shown i n Fig. 9 adjacent to corresponding mi- c r o s t r u c t u r e s . zones, but not with the s a m e degree of delineation as the be t a -gamma autoradiographs. m i c r o s t r u c t u r e and the movement of f iss ion products .

The alpha autoradiographs somet imes showed the annular

The autoradiographs c l ea r ly indicate a re la t ion between the

I. Examination of Separa ted Annular Zones

Four t r a n s v e r s e sect ions f r o m near the middle of the fuel columns of ND28, 30, 35, and 43 w e r e s t r ipped of the i r cladding to p e r m i t a m o r e detai led s tudy of the annular zones. eas i ly separa ted mechanical ly with a s imple punch and die. The center sec t ion general ly remained intact , but the middle and outer zones f r ag - mented.

The t h r e e ma jo r annular zones w e r e

F igu re 18 shows the typical appearance of the sepa ra t ed zones.

Density measu remen t s w e r e made for the annular sect ions f r o m ND30, 35, and 43 by i m m e r s i o n in toluene. men t s are shown in Table XVII. dense (7.5-9.4 g /cm3) and the middle zone consis tent ly m o s t dense (14.4- 16 .2 g/cm3). or iginal fuel.

The r e su l t s of these m e a s u r e - The center zone is consis tent ly l e a s t

Two of the densi ty values fo r ND43 a r e g r e a t e r than the

The annular zones f r o m ND28, 30, 35, and 43 w e r e analyzed fo r uran ium, plutonium, z i rconium, technet ium, and a l so , i n the c a s e of ND30, for neodymium. to de t e rmine i f material f r o m the cladding had penetrated into the fuel at the in te r face . isotopic-dilution mass spec t romet ry , with an ove ra l l e r r o r es t imated at 5- 10%. es t imated e r r o r of 207’0, and technet ium was m e a s u r e d spec t rochemica l ly with a n es t imated e r r o r of 5%. was done by isotopic-dilution mass spec t romet ry .

An analysis f o r nickel i n the zones f r o m ND35 was made

The u ran ium and plutonium contents w e r e de te rmined by

-. Zirconium content was de te rmined spec t rographica l ly with a n

The one s e t of neodymium-148 ana lyses The r e su l t s of t hese

30

v) z 0

F: t: v)

D

W I

v)

-I 0

v)

-

a

a

a

w

m

m

0

'f' co 0

c)

4d

d

z a a, 4

t n

d

.rl

cd X c a

oil iL"

31

FP-7090 lox

(a) Element ND28, bottom

FP-8383 lox

(c) Element ND32, bottom

FP6993 lox (e) Element ND43, bottom

Note: A l l surfaces as-polished.

FP-7097 lox

(b) Element ND28, top

FP-8395 lox

(d) Element ND32, top

FP-7004 lox

( f ) Element ND43, top

Fig. 10. End Effects in Fuel Columns

FP-7100 15X

(a) ND30 (U-14Pu-lZZr)

FP-6996 15X

(b) ND43 (U-15Pu-9Zr)

Fig. 11. Separations in Fuel Columns, As-polished

33

FP-8282 250X

(a) Element ND32, Longitudinal Section, Center Zone at Midlength of Fuel Column

FP-6591 250X

(c) Element ND43, Longitudinal Section, Boundary between Center and Middle Zone a t Midlength of F u e l Column

FP-6942 250X

(b) Element ND43, Transverse Section, Boundary between Center and Middle Zones a t Midlength of Fuel Column

Fig. 12

Variations in Porosity of U-15Pu-9Zr Fuel, As -polished

34

FP-6610 250X

(a) Element ND43, Transverse Section, between Middle and Outer Zone

FP-6029 250X

(b) Element ND28, Longitudinal Section, Middle Zone

Fig. 13. Internal Cracking in U-15Pu-9Zr Fuel, As-polished

h

v

Iu

z! 9q

r.

5 c I V

c

I

-I N

N

3

(D

n

-0

I a

a3 a

P

E

m

W

W a

N

VI x"

w

cn

36

Fig. 16

Banding at Boundary between Middle and Center Zones of Element ND35 (U-14Pu- 12Zr). Transverse section near top of fuel column; etched with ammonium phosphate.

FP-8149 250X

Fig. 1 7

Cellular Microstructure in Element ND35 (U-14Pu- 12Zr). Transverse section, middle zone about 3 in. from bottom of fuel, as- polished .

FP-8071 250X

37

A

FP-8080 FP-8078

(a) Center Zone

7x

. . . . . .

. I . . ::.: '. ;. . .

.>*.

(b) Middle Zone

7x

Fig. 18

Separated Annular Zones of Element ND30 (U-14Pu-12Zr)

FP-8083 7 x (c) Outer Zone

T A B L E XVII. D e n s i t i e s of A n n u l a r F u e l S e c t i o n s

O r i g i n a l F u e l

F u e l D e n s i t y P o s t i r r a d i a t i o n D e n s i t i e s (g/ c m 3 )

E l e m e n t (g / cm3 1 C e n t e r Z o n e Midd le Zone O u t e r Z o n e

ND30

ND35

ND43

1 5 . 3

15 .3

15 .8

9 .1 $:; (3 )a

9 . 4 ( 3 )

7 . 5 ti:; ( 3 )

a N u m b e r s i n p a r e n t h e s e s are t h e n u m b e r of d e t e r m i n a t i o n s on s a m p l e .

38

ana lyses , except fo r neodymium and nickel, a r e given in F ig . 19. dymium resu l t s i n t e r m s of the a tomic ra t io Nd/(U t P u ) , based on the or ig-

The neo-

inal composition, w e r e : cen ter zone,

ou te r , 8.8 x u r ed s pe ct r ophot ome t ri cally with the following r e s u l t s : cen ter zone,

\o 6.8 x middle , 3 .3 x and Nickel was m e a s - i

E 30 2 5

5 20

u none detected; middle , none detected; I10 2 5 ou te r , 4.4 5 0.20Jo. 8 E O

8 15

ANNULAR ZONES J. Burnup Analyses ANNULAR ZONES e

Burnup samples f r o m the en- t i r e fuel-element sec t ion w e r e taken f r o m the top, middle , and bottom of the fuel columns of ND32,43, 30, and 35. The burnup r e su l t s obtained by techne-

!i CENTER MIDDLE OUTER ANNULAR ZONES ANNULAR ZONES

J 0 ND-28 t ium analysis a r e given i n Table XVIII. ~ ND.43}9w/oZr iEI:g}12w/oZr

The four e lements w e r e se lec ted as represent ing two spec imens each of U-15Pu-9Zr and U-14Pu-12Zr .

Fig. 19. Compositional Vari%ons in Annular Zones of U-Pu-Zr Fuel Pins

TABLE XVIII. Burnups by Technet ium Analysis

Ca lc Burnupa Bur nupa by Analysis

Element Max Avg Top Middle Bottom ~ ~~

ND32 4.2 3.7 2.4 4.0 3.2 ND 43 4.3 3.8 2.9 3.9 3.2 ND30 4.7 4.1 3.0 4.4 3.5 ND35 4.5 4.0 3.1 4.1 3 .5

aOf heavy a toms.

K. Metallography and Microprobe Analysis of the Cladding

F ive fuel-cladding combinations w e r e examined e i ther by metal lography o r by e lec t ron-microprobe analysis to de t e rmine the extent of fuel- cladding interact ion. The combinations examined with the mic roprobe w e r e U - 15Pu-9Zr clad with Type 304L s ta in less s t ee l (ND28), U - 14Pu- 12Zr clad with Type 316 s t a in l e s s steel (ND30), and U-14Pu-12Zr clad with Hastel loy-X (ND35). Combinations of U-15Pu-9Zr with Hastelloy-X (ND43) and U-15Pu- 9 Z r with Type 3 16 s t a in l e s s s t ee l (ND32) w e r e examined metal lographical ly . 8

Nonuniform reac t ions between the fuels and the claddings w e r e evident i n the meta l lographic examinat ions. max imum th ickness , was found in Type 316 s t a in l e s s s t ee l ; no apparent dif- f e r ence exis ted between 9 w/o Z r and 12 w/o Z r alloys clad with e i ther Type 316 s t a in l e s s steel o r Hastel loy-X. The depth of r eac t ion i n Type 304 s t a in l e s s steel and Hastel loy-X w e r e s i m i l a r , with max imum th icknesses of 0.005 in. However , the r eac t ion l aye r in Hastel loy-X was m o r e uni form. The reac t ions genera l ly o c c u r r e d i n the middle and upper port ions of the e lements , w h e r e fuel- cladding in te r face t e m p e r a t u r e s ranged f r o m 530 to 650°C.

The l e a s t reac t ion , of 0.001 -in.

Typical fuel-cladding reac t ions a r e shown in Fig. 20. F i g u r e 20(c) shows a c r a c k between the reac t ion l a y e r and the una l te red Hastelloy-X. The p r e s e n c e of the c r a c k impl ies s o m e degree of b r i t t l eness in the r e a c - t ion l a y e r , and could be the r e su l t of a difference i n the coefficients of t h e r m a l expansion of the r eac t ed and unreac ted m a t e r i a l s . One sec t ion of ND43 showed a shea r - type c r a c k in the cladding [ s e e F ig . 21(a)]. The c r a c k begins at the outs ide su r face and t e rmina te s at the edge of the r e a c - t ion zone. appeared to be incipient c r a c k s on the ins ide of the cladding [ s e e Fig. 2 l (b) ] .

A sec t ion of Type 316 s t a in l e s s s t ee l (e lement ND30) had what

Cladding spec imens of Type 304L and 316 s t a in l e s s s t ee l , and Hastel loy-X w e r e taken f r o m n e a r the midlength of the fuel column, sepa - r a t e d f r o m the fuel , and examined in g r e a t e r detail .

A mounted and polished spec imen of Type 304L s t a in l e s s s t ee l f r o m e lement ND28 ( 9 Z r ) was sa t i s fac tor i ly etched electrolyt ical ly with 2Q70 oxalic acid [ s e e F ig . 22(a)]. The s t r u c t u r e of the unreac ted metal is c l ea r ly shown, but the s t r u c t u r e of the r eac t ion zone is not brought out. The i r r e g u - l a r i t y i n the depth of the r eac t ion zone is evident. dence that the reaction zone p r e f e r s to follow g ra in boundar ies , Aprec ip i t a t e (presumably ca rb ide ) is p r e s e n t i n the g ra in boundaries and i n the g ra ins . This prec ip i ta te is similar to tha t obse rved i n uni r rad ia ted tubing heated at 600°C f o r 96 h r [see Fig . 23(a)]. In the uni r rad ia ted , hea t - t r ea t ed tubing t h e r e is a m a r k e d carb ide precipi ta t ion n e a r the ins ide wall and a lesser precipi ta t ion throughout the r e m a i n d e r of the wall. The local ized prec ip i - ta t ion toward the ins ide wal l p re sumab ly was due to the 770 cold work p r o - duced dur ing the p r o c e s s of tube expansion. s t a in l e s s steel cladding was m o r e magnet ic than the hea t - t r ea t ed un i r r ad i - a ted cladding.

T h e r e is no spec i f ic evi-

Also, the i r r a d i a t e d Type 304L

39

The i r r a d i a t e d Type 316 s t a in l e s s s t e e l cladding [see Fig. 22(b)] had a prec ip i ta te phase i n the g r a i n boundaries and a f ine par t icu la te phase within the g ra ins . Both of t hese w e r e p re sumab ly carb ides . the r eac t ion zone is not un i form, but grain-boundary penetrat ion was not observed . In a n uni r rad ia ted sample hea t t r ea t ed for 96 hr at 600°C

The depth of

[ s e e Fig. 23(b)], the g ra in s i z e and the grain-boundary ca rb ides w e r e e s sen - t ia l ly ident ical with those of the i r r a d i a t e d m a t e r i a l , but t h e r e was no fine par t iculate phas e within the g ra ins .

EI-6112 250X

(a) Element ND28 (9Zr), Type 304L Stainless Steel Cladding (top), Transverse Section

EI-6136 250X

(b) Element ND30 (12Zr). Type 316 Stainless Steel Cladding (top), Transverse Section

Fig. 20

Fuel C ladding Interactions, As -polished

FP-6858 250X

(c) Element ND43 (9Zr), Hastelloy-X Clad- ding (top), Longitudinal Section

Crack

.

FP -6588 50X

(a) Element ND43, Hastelloy-X Cladding, Transverse Section

Fissures

.

Q

E14135 250X

(b) Element ND30, Type 316 Stainless Steel Cladding, Transverse Section (incipient cracks -1/2 mil deep)

Fig. 21. Cladding Cracks

41

.

47 5 16 500X

(a) Element ND28, Type 304L Stainless Steel Cladding, Electrolytically Etched with 2-1/27”. Oxalic A c i d

48733 500X (b) Element ND30, Type 316 Stainless Steel

Cladding, Electrolytically Etched with 2-1/2770 Chromic A c i d

48 7 43 500X

(c) Element ND35, Hastelloy-X Cladding, As -polished

Structures of

Fig. 22

lrra di ate :d Claddings

.

4

.

47518 500X

(a) Type 304L Stainless Steel, 6OO0C, 96 hr, Electrolytically Etched with 2 -1/270 Chromic Acid

48132 500X

(b) Type 316 Stainless Steel, 6OO0C, 96 hr, Electrolytically Etched with 2-1/2$~ Chromic A c i d

.

Fig. 23

Structures of Heat-treated Claddings

48741 500X

(c) Hastelloy-X, 6OO0C, 96 hr, Elec- trolytically Etched with 2-1/270 Chromic A c i d

44

The s t ruc tu re of the i r r ad ia t ed Hastelloy-X was not de te rmined be- a cause of etching p rob lems [see F ig . 22(c)]. The uni r rad ia ted HastelIoy-X spec imen [see F ig . 23(c) ] that was heated at 600°C f o r 96 h r appeared to have a no rma l s t ruc tu re , but had a g ra in s ize l a r g e r than that in e i ther of the types of s t a in l e s s s teel .

Cladding spec imens of Types 304L and 316 s ta in less s teel , and Hastelloy-X, taken f r o m nea r the midlengths of the fuel columns of ere- ments ND28, 30, and 35, respect ively, were examined with an electron microprobe . cladding.

The scans included both the reac t ion zone and the unreacted Typical r e s u l t s (for Type 316 s ta in less s tee l ) a r e shown in F i g . 24.

The max imum penetrat ion of the reac t ion zone into the specimen

(ND28) of Type 304L s ta in less s t ee l was 140 pm (5.5 mi ls ) . products (La , Ce, Pr, Nd, and Sm) w e r e detected in the reac t ion zones but nei ther uran ium nor plutonium was found. affected cladding and the reac t ion zones the nickel content dec reased sharply f r o m 10 to 1%. The manganese content of the reac t ion zone was the same a s that of the unaffected steel. I ron and ch romium concentrat ions inc reased slightly, and sil icon concentrat ion dec reased slightly in the reactJon zQne as compared with the unaffected cladding.

Solid f iss ion

At the boundary between the un-

Type 316 s ta in less s t ee l cladding f r o m ND30 (:!.270 Z r ) was si.mi.Farly examined with the e lec t ron microprobe . The reac t ion zone was i r regul .ar i n thickness and only f r o m 25- to 30-pm (-1 mil) thick at its g r e a t e s t penetra... tion. Type 316 s ta in less s t ee l ciadding as w e r e found in Type 304L stai.nl.ess s tee l ( s e e F ig . 24). In addition, both uran ium and plutonium w e r e identified i.n the reac t ion zone. dropped f r o m 13 to about 1 w/o. The i r o n concentrat ion in the react ion ~ o r x dec reased , but the ch romium inc reaseds par t icu lar ly near the fue1..-cladding in te r face . The 0veral . l si l icon concentration dropped, but t.he-Pe was a s m a l l peak at the in te r face . molybdenum inc reased slightly at the interface. The microprobe ana lyses for the meta l l ic e lements indicated no r eason for the thinner reac t ion zone i n the Type 316 stai ,nless s t ee l as compared with both the Type 304L s t a i m . - l e s s s t ee l and the Hastelloy-X,

The same solid fi.ssion products were found in the reac t ion zone of

The nickel content of the cladding in the reac t ion zone had

The manganese concentrati.on was unaifec.ted, and

The maximum penetrat ion of the reac t ion zone into the HasteIloy-X cladding of ND35 (12Zr) was 152 p m ( 6 mi l s ) . r e s u l t s for Hastelloy-X differed f r o m those f o r the types of stajnit-ss steel i n that, of the solid f i ss ion products , only c e r i u m and neodymium were d e - tected. Chromium and, to a l e s s e r extent, i r o n were somewhat m o r e con- - cent ra ted in the reac t ion zone than in the unaffected cladding. concentrat ion i n the reac t ion zone was 15 W/O compared with 9 W/O In the

The microprobe-aca lys i s

Molybdenum

0

Q 90

60

30

0

I

I

u C a, L) Ll a, PI

u 6 *d

I 6

4 , - 1 - I Mn

6 -

4 ,

1

I

I Mo

-

t- 4-1 Reaction I Nominal

ype 316 Stainles

‘Cladding - Fuel

6000

4000

2000

0 6000

4000

2000

0

600

400 a G

al m

$ 200

Ll 0

* 1200 C 800

400

0 600

400

al

v) u

U

200

0

1200

800

400

0

1

Pu

I I U

I l*l Pr

6oo 400 1-1 Interface 0

Fig. 24. Typical Distribution of Elements in Surface Layer in Irradiated Type 316 Stainless Steel Cladding (U-Pu-Zr Fuel)

or ig ina l meta l . t ion zone than i n the unaffected cladding. p re sen t i n the reac t ion zone adjacent to the in te r face .

Sil icon concentrat ion a l so tended to be h igher in the r e a c - Both u ran ium and plutonium w e r e

L. Hardness T e s t s on Cladding

Hardness t e s t s w e r e m a d e on two of the i r r a d i a t e d cladding ma te - Type 304L s t a in l e s s s t ee l (0.020 in. thick) f r o m ND28, and Type 316

F o u r DPH h a r d n e s s m e a s u r e - rials: s t a in l e s s s t ee l (0.015 in. thick) f r o m ND30. men t s with a 0.5-kg load w e r e m a d e i n the r eac t ion zone on a t r a n s v e r s e sec t ion of Type 304L s t a in l e s s s t e e l f r o m near the midlength of the fuel. The r e su l t s w e r e 281 ?;I DPH. Eighteen h a r d n e s s m e a s u r e m e n t s outside the react ion zone gave 208 DPH. Two sets of five m e a s u r e m e n t s each w e r e made on a longitudinal sec t ion of Type 316 s t a in l e s s s t ee l about one-third of the d is tance f r o m the bot tom of the fuel column. The small reac t ion zone was not included i n the h a r d n e s s m e a s u r e m e n t s on Type 316 s t a in l e s s steel. at the ins ide : Set 1 , 258 compared with 215, ave rage 238 ff!; Se t 2 , 270 compared with 235, ave rage 258 n e s s e s of 150-180 i n the annealed and uni r rad ia ted condition.

In both sets, the h a r d n e s s at the outs ide was higher than

DPH. Both materials had DPH h a r d -

VII. DISCUSSION

A . Dimensional Changes of the Fue l Elements

.

Changes i n both the ex terna l dimensions and the geomet ry of the fuel e lements have been shown by m e a s u r e m e n t s of bowing, length, d i am- e t e r , vo lume, and weight after i r rad ia t ion . The i n c r e a s e s noted in bowing a re probably caused by the t h e r m a l s t r e s s e s genera ted during r e a c t o r operat ion. s t r e s s e s a r i s ing f r o m the tube expansion p r o c e s s (7% cold work) could a l s o have been a fac tor . because the d i rec t ion of bowing while i n the r e a c t o r was not identifiable,

However , for the Type 304 s t a in l e s s s t ee l cladding, re l ief of

Evaluation of the change i n bowing is not poss ib le

All of the 14 fuel e lements examined showed length i n c r e a s e s ranging f r o m 8 to 44 mils ( s e e Table XII). T h e s e m e a s u r e m e n t s would be affected by the bowing of the fuel e lements . I f the bowed e lements w e r e cons idered as arcs of c i r c l e s , s t ra ightening of the g r e a t e s t bow (0.28 in . ) would add 7 mils to the length; however , the ave rage bow would only con- t r ibu te about 2 mils to the length i n c r e a s e of each element .

Estimates of s o m e of the s t r a i n s that would contr ibute to the m e a s - u r e d length i n c r e a s e s of the fuel e lements follow. A pos t i r rad ia t ion cladding t e m p e r a t u r e higher by 1°C than the p re i r r ad ia t ion t e m p e r a t u r e would produce a length i n c r e a s e of 0.5 mil due to thermal expansion. However, t h e r e was no evidence that decay hea t (less than 10 W/pin) caused a cladding

t e m p e r a t u r e m o r e than 2 o r 3°C above ambient. T e m p e r a t u r e differences have been e s t ima ted to account fo r 1-2 mils of the length i n c r e a s e s of t he fuel e lements . The calculated gas p r e s s u r e i n the plenums ( s e e Table XI) at the end of i r r ad ia t ion would produce an e l a s t i c s t r a i n in the cladding of about 3 mils over the length of the e lements . The axial component of the volume change f r o m fas t -neut ron damage would contr ibute about 55 mils to the pos t i r rad ia t ion 1ength.l9e2’ This last fac tor is the l a r g e s t of those evaluated, and is g r e a t e r than the l a r g e s t m e a s u r e d length i n c r e a s e . ra tchet ing of the fuel and cladding af te r contact is apparent ly ru led out by the fac t that two fuel pins sh rank axially 0.7 and 1.07’0, whereas the c o r r e - sponding e lements i n c r e a s e d in length by 38 and 10 m i l s , respec t ive ly . difference between the two fuel compositions apparent ly was not impor tan t i n the length i n c r e a s e s of the fuel e lements .

t

s

Axial

The

. -

A definite co r re l a t ion between length i n c r e a s e s and cladding com- posit ion is not obvious. 34, and 44) had the l a r g e s t length i n c r e a s e s , but the o ther t h r e e (ND35, 37, and 39) had among the smallest length i n c r e a s e s . The Type 304 s t a in l e s s s t e e l cladding s e e m e d to have lengthened slightly m o r e than the Type 316 s t a in l e s s s tee l .

T h r e e of the nickel-base alloy claddings (ND43,

Although the evidence fo r d i ame t ra l changes of the fuel e lements i s marg ina l because of the l imi ted prec is ion of the m e a s u r e m e n t technique, the d i a m e t e r s i n the sect ion occupied by the fuel showed s o m e evidence of i n c r e a s e . The r e s u l t s have been s u m m a r i z e d i n Table XII. No tendency toward ovali ty was shown. Individual d i a m e t r a l i n c r e a s e s of as much as 1 mil w e r e m e a s u r e d . genera l ly along the upper half of the fuel column, of a prof i le of d i ame t ra l i n c r e a s e s that could be at t r ibuted to cladding swelling induced by fas t -neut ron exposure (max imum fluence of 3 . 4 x lo2‘ nvt). to a s t r a i n of 0.570~ r e l a t ed with e i ther fuel o r cladding composition.

The l a r g e r individual d i a m e t r a l i n c r e a s e s w e r e T h e r e was no indication

A 1-mi l -d i ame t ra l change i n these elements would be equivalent The d i a m e t r a l i n c r e a s e s cannot be conclusively c o r -

The volumes of the fuel e lements all i n c r e a s e d slightly as a r e s u l t of the r e a c t o r exposure ( s e e Table XIII). less s tee l -c lad e lements showed the g r e a t e s t vo lume i n c r e a s e (0.25% ave rage ) , followed by the e l emen t s c lad with Hastelloy-X-280 (0.20% ave rage ) . Hastel loy-X cladding, showed ave rage volume i n c r e a s e s of 0 .14 and 0.13q;’o, respect ively. vo lume i n c r e a s e s .

showed apprec iab le differences in volume i n c r e a s e s , the i n c r e a s e s m u s t b e caused by f ac to r s other than the cladding composition. A re la t ionship has been found between the volume i n c r e a s e and the geomet ry of the

As a group, the Type 304L s ta in-

The o ther two g roupss with Type 316 s t a in l e s s s t e e l and

The fuel composition does not seem to b e re la ted to these Also , because fuel e lements with similar claddings . (Types 304 and 316 s t a in l e s s steel; Hastel loy-X and Hastelloy-X-280)

cladding, a s shown in Fig. 25: the g r e a t e r the cladding volume pe r unit length ( r ep resen ted by the c ros s - sec t iona l a r e a of the cladding), the

0.3

.\" w v) 0.2 a a W

u z W I 3 J

>

-

0 0.1

n

0.0. I. 0. 0.206 0.166 0.196 0.166

J TYPE 334 S S 3 TYPE 316 SS 1 HASTELLOY-X 0.196 0.166 > HASTELLOY-X-280 0.206 0.1 76

--

g r e a t e r the volume i n c r e a s e of the element , Since all the e lements a r e the s a m e length, the c ros s - sec t iona l a r e a s of the claddings a r e a l so representa t ive of the i r total volumes,

Two fac tors that can cause a volume change i n the e lements and that a r e a l so r e - la ted to the c ros s - sec t iona l a r e a , o r volume, of the cladding a r e the rma l ra tchet ing and cladding swelling. The calculated the rma l and p r e s s u r e s t r e s s e s ( s e e Table XI) indicate that t h e r m a l ra tchet ing i s a possibi l i ty , based on the Mil ler c r i t e r i a fo r growth, only during the f i r s t t he rma l cycle. A calculation of the effect of one the rma l cycle on the length and

" 8 9 IO I I 12 d iame te r of the element gives values that

306-197 Rev. 2 measu red . Neither the length nor the d iame-

Fig. 25. Av of Fuel Element vs Cross Section of Cladding s ion of the e lements . This behavior suggests

CROSS SECTION OF CLADDING, 10-'in' a r e of the s a m e o r d e r of magnitude as those

t r a l measu remen t s indicate i so t ropic expan-

that cladding swelling is not the dominant mechan i sm for the volume changes. cladding swelling as functions of fluence and t empera tu re for the volume ave rage t e m p e r a t u r e of the claddings (fluences of 3.9-3.4 x nvt) indi- ca te that these claddings should have swelled an average of 0.570. This value i s g r e a t e r by about a fac tor of two than the measu red volume in- c r e a s e s of the fuel e lements . ca l values and the m e a s u r e d changes cannot be reconci led a t this t ime.

Quantitative relationships '9120 fo r

The lack of cor re la t ion between the theore t i -

All of the fuel e lements experienced weight l o s s e s pe r unit su r f ace a r e a between 6.5 and 11.5 mg/dm2. T h e r e was no indication that these weight l o s s e s w e r e due to f iss ion-gas r e l ease . The max imum weight loss is equivalent to a reduct ion of 4.9 x These weight l o s s e s occur red in s ta t ic sodium that contained f r o m 30 to 50 ppm of oxygen a t the s t a r t of i r rad ia t ion . nificant. to be at l e a s t a factor of ten g r e a t e r , depending on specif ic conditions.

B.

i n . /y r in the cladding thickness .

They a r e not considered s ig - The weight l o s s in flowing sodium in a r eac to r would be expected

Fuel Swelling and F i s s ion -gas Re lease

The i n c r e a s e s in fuel volume w e r e l a r g e r i n the elements with 66% s m e a r densi ty than i n those with 7570 s m e a r density. ca tes that in the low-density e lements the fuel swelled m o r e before contact- ing the cladding and that the cladding r e s t r a ined the swelling af ter contact. The r e s t r a i n t apparent ly did not produce significant s t r a i n i n the cladding.

This behavior indi-

49

Nei ther the burnup at which fuel-cladding contact o c c u r r e d , nor the swel l - ing r a t e before o r after contact can be de te rmined f r o m these spec imens .

A c

i

The var ious contributions to fuel swelling a r e es t imated as follows. A portion of the volume i n c r e a s e of each fuel column equivalent to 9-1070 of the original fuel volume is attr ibuted to the formation of solid f iss ion prod- uc ts . Table XV) . The ave rage burnups f r o m Table VI11 w e r e used in these calcu- la t ions. 4670 of the or iginal fuel volume, is attr ibuted to the f iss ion gases xenon and krypton and to anisotropic growth of the fuel, voids and a swir led effect in the fuel f r o m nea r the bottom and adjacent to the cladding of ND35. This location is a region of re la t ively low t empera - t u r e , f r o m 475 to 600°C ( s e e Fig. 4). The material has the appearance of mechanical tear ing in alpha uranium as descr ibed by Leggett et aJ~321 and by Angerman and Caskey.22 The overa l l extent of the swelling assoc ia ted with the tear ing cannot be evaluated f r o m our p re sen t knowledge of these fuel e lements

This value is based on 2+70 swelling p e r a tomic percent burnup ( s e e

The remaining portion of each volume i n c r e a s e , equivalent to 2 2 -

F igu re 15 shows i r r e g u l a r

F i s s ion -gas generat ion has been calculated f r o m the average calculated burnup for each fuel e lement on the bas i s of 0 . 2 7 a tom of gas for each a tom fissioned. The max imum p r e s s u r e s and corresponding s t r e s s e s that could develop as a r e su l t of gas production are given in Table XI. The s t r e s s e s a re well below the s m a l l e s t c r eep - rup tu re design s t r e s s (10,000 p s i for 10,000-hr l ife at 675°C) for these cladding m a t e r i a l s . f iss ion gases a re retained in the fuel and a r e prevented f r o m expanding by the s t rength of the fuel, by the su r face tension, and by the r e s t r a in ing fo rce of the cladding. The f r e e fuel sur faces consis t of both m a c r o - and m i c r o - c r a c k s ( s e e F igs . 11 and 13) and of interconnected porosi ty , all of which p e r m i t the evolved gases to pas s to the plenum, In the eight e lements f r o m which gases were collected, the r ecove ry of f i s s ion gases ( s ee Table X V I ) ranged f r o m 60 to 7170 of the quantity calculated f r o m the ave rage burnup. No difference i n the r e l e a s e of f iss ion gases f r o m the U-15Pu-9Zr fuel as compared with U-14Pu-12Zr is indicated. has shown a c o r r e l a - t ion between f iss ion-gas r e l e a s e and fuel swelling for U - F s and U - P u - F s . As shown in Fig. 8 , the da ta f r o m the eight U - P u - Z r e lements a g r e e with his cor re la t ion .

Actually, s o m e

The plenum original ly contained he l ium g a s , nominally at a glovebox p r e s s u r e of -1 /2 in. HzO at r o o m tempera tu re . seven e lements was f r o m 76 to 9170 of that expected. yielded only 4070 of the expected amount of hel ium. i n Table XVI. suspicion on the r e su l t s for he l ium recovery . s u r e possibly could have forced s o m e hel ium out by gas expansion before the weld was completed. men t is difficult to account for by this p rocess .

The r ecove ry of hel ium f r o m An eighth e lement

T h e s e r e su l t s a re shown 0 T h e r e was nothing in the overa l l f iss ion-gas r e su l t s to cast

Heat f r o m the final weld clo-

However, a 60% loss of he l ium f r o m the one e le -

An a t tempt was m a d e to calculate plenum volumes after i r r ad ia t ion by backfilling six e lements with nitrogen. to calculate final plenum p r e s s u r e s on the b a s i s of the amounts of gas col- l ec ted . (average 80700) of the p r e s s u r e s shown in Table XI. between the two methods i s only f a i r , it conf i rms that the plenum p r e s s u r e s would not have l imi ted fu r the r i r r ad ia t ion of t hese e lements .

The values obtained w e r e used

The values ranged f r o m 273 to 625 psi ; they a r e f r o m 51 to 10970 Although the ag reemen t *

C. Fuel-pin Behavior

Typical separa t ions in the fuel column are shown in Fig. 11, In Fig. 11 ( a ) , the separa t ion probably developed e i the r in e a r l y o r midlife of the element , because the middle annular zone h a s moved p a r t way into the wide c r a c k i n the outer zone. oped at the end of life of the e lement , because no obvious d is tor t ion of mat ing su r faces occur red . However , whether the wide sepa ra t ion i n the outer zone r e p r e s e n t s m a t e r i a l pulled out during the grinding operat ion o r whether the sepa ra t ion is a shift of the outer zone re la t ive to the middle and center zones is not certain.

The separa t ion in Fig. 1 l ( b ) probably devel-

A possible explanation of the axial sepa ra t ions , under the assumption that the three annular zones a r e assoc ia ted with the phase t ransformat ions of the U - P u - Z r s y s t e m , is the different ia l t h e r m a l expansion that occur s in the fuel. dur ing r eac to r s t a r tups and shutdowns. The t h e r m a l expansion coefficients of the alloy as a function of t e m p e r a t u r e a r e 5

25-595°C 17.6 x in./in.-"C

In t r ans fo rma t ion r ange 5.2 x in./ in.

665--7 50°C 20.1 x i n . / i n . - " ~

Application of t hese coefficients to the 13-in0-long fuel column, with the assumpt ion of un i form axial t e m p e r a t u r e s , gives a different ia l expansion of 0 .068 in. between the unt ransformed m a t e r i a l at 595°C and the fully t r ans fo rmed alloy at 665OC. The different ia l expansion fo r the fully t r a n s - fo rmed alloy between 665°C and -750°C ( the max imum center t e m p e r a t u r e ) is 0,023 in. fo r the 13-in0-long fuel column, s ion of the center of the fuel column re la t ive to the outs ide zone is 0.091 in , , o r 0,770. 25"CO8 Similar tests at 675°C produced ductile f r ac tu re . Thus, different ia l t h e r m a l expansion alone could be respons ib le for the separa t ions i n the fuel columns

The total different ia l expan-

Alloys of U - P u - Z r b r e a k i n a b r i t t l e fashion in . t ens i l e t e s t s at

. Axial growth of the fuel s e e m s to be re la ted to t empera tu re .

co r re l a t ion is shown in Fig. 26, where the change in fuel length ( s e e Table XII) is plotted v e r s u s the volume ave rage fuel t e m p e r a t u r e (see Table X).

This

The higher f i s s i l e content of the 970 Z r fuel compared with the

51

TYPE 304

A HASTELLOY X

0 TYPE 3 0 4

0 HASTELLOY x-zso

0

+ 2 -

+I -

0 -

-I I I I I 580 590 600 610 620 630 640

AVERAGE TEMPERATURE OF FUEL VOLUME ,'C

306-198 Rev. 1

Fig. 26. Length Change vs Tem- perature of Fuel

12770 Z r e lements appears sufficient to account for the i r posit ions at the high- t empera tu re end of the graph . The da ta for the elements with 65% smear density a l so a g r e e with this gene ra l t rend . The g rea t e r axial growth at the lower t e m - pe ra tu res might a l so explain the preva- lence of separa t ions in the lower half of the fuel column. The tendency for g r e a t e r growth of the fuel column where t empera - t u r e s a r e lower would be expected to r e - sul t in the generat ion of m o r e stress.

The two fuel pins (ND35 and 43) that showed no axial separa t ions w e r e also the two that dec reased in length. Both had high volume-average t e m p e r a - t u r e s of -620°C (calculated without con- s ider ing t empera tu re i n c r e a s e s because of the reduced the rma l conductivity due to pore formation) . The likelihood is

v e r y low that the fuel columns separa ted and then joined again without giving s o m e indication on the neutron radiographs. probably never elongated. A reduction in swelling r a t e at the higher tem- p e r a t u r e would not account for a d e c r e a s e in length. weight of the fuel column at the higher t empera tu re is a m o r e likely explanation. However, the fact that some fuel pins a t higher calculated tem- pe ra tu res did not d e c r e a s e in length indicates that an adequate explanation for length changes is lacking.

Thus, t hese columns

C r e e p due to the

D. Fuel S t ruc tu re

The annular zones observed on the as -polished metal lographic sect ions ( s e e Fig. 9 ) of ND35 are typical of the five fuel e lements examined. The observed s t r u c t u r e s of the fuel columns did not seem to have been affected by the differences in fuel composition ( 9 Z r v e r s u s 12Zr ) . t he bottom ends of the fuel columns the annular zones w e r e not developed, but they w e r e well developed toward the upper ends. t h r e e quite sharp ly delineated m a j o r zones. different f r o m the regions inside the zones and constituted minor zones. The center zone appeared m o s t porous, and the middle zone l e a s t porous. Some of the neutron radiographs indicated the differences in porosi ty as a discontinuous var ia t ion in film densi ty a c r o s s the d i ame te r of the fuel. In s o m e cases the annular regions w e r e d is tor ted , probably as the r e s u l t of loca l t empera tu re effects.

Toward

T h e r e were usual ly The b o r d e r s between appeared

The bottom ends of the fuel columns had a different appearance than the top ends ( s e e Fig. 10). At the bottom a conical sect ion tended to s e p a r a t e

f r o m the center of the fuel column. At the top the en t i r e end tended to be pushed off. s ions on the effectively unres t ra ined ends a r e (1) differential t he rma l expansion of the phases in the fuel column; ( 2 ) f iss ion-gas p r e s s u r e in the po res of the cen t r a l zone; and (3) compress ive fo rces f r o m the swelling of the outer zone. All of t hese fac tors w e r e probably operat ive, but to different deg rees .

Among the possible s o u r c e s of the dr iving f o r c e fo r the extru-

Longitudinal sect ions ( s e e Figs . 9 and 10) showed that, i n genera l , the zones tapered off rather rapidly at the ex t r eme top of the fuel column. At the bottom the re was a much m o r e gradual taper ing of the zones, which seemed to t e rmina te s o m e dis tance away f r o m the bottom end. End effects w e r e not cons idered in the t empera tu re calculations. ance of the top ends of the fuel columns indicated that the sodium above the fuel column s e r v e d as a good medium for heat t r a n s f e r f r o m the fuel to the cool ant

However , the appear -

The center zone tended to be uniformly porous , except toward the edge, where a nar row, re la t ively nonporous band was somet imes encoun- t e r e d [ s e e F igs . 9 , 10, and l l ( b ) ] . the middle zones was usual ly well defined. s epa ra t ed by a c rack . did have s o m e porosi ty . common i n the middle zone ( s e e F igs . 11 and 13). The middle zone in s o m e locations had a r e a s of small po res outlined by l a r g e r po res ( s e e Fig. 17), the overa l l appearance being somewhat similar to the columnar g r a i n region in oxide fuels. the m i c r o s t r u c t u r e in Fig. 16. bottom of the fuel column, a f te r the center zone d isappeared , and blended with the outer zone. The outer zone was quite porous9 but the po res w e r e genera l ly smaller than i n the center section. r ep resen ted gas bubbles o r pull-outs f r o m polishing, w e r e encountered in the outer zone. appearance than the bulk of the outer zone, presumably because of the fuel- cladding react ion.

The boundary between the center and Somet imes the two zones w e r e

The middle zone appeared to b e m o s t dense , but it Interconnected porosi ty and m i c r o c r a c k s w e r e

Banding i n the middle zone is i l lus t ra ted by This middle zone continued toward the

L a r g e voids, which m a y have

The fuel reg ion near the cladding usually had a different

Atypical m i c r o s t r u c t u r e s observed on as-pol ished spec imens a r e shown in F igs . 14(a) and 14(b). needles of oxygen-r ich alpha z i rconium on the boundary between the center and middle zones. F igu re 14(b) shows a m i c r o s t r u c t u r a l consti tuent on the inside edge of the outer zone. the middle zone to the outer zone, as d iscussed below.

F igu re 14(a) shows what appear to b e

This consti tuent was p re sen t in the uni r rad ia ted m a t e r i a l

This m a y b e UZr, produced by movement of z i rconium f r o m

The middle and the outer zones commonly broke up when sepa ra t ed for densi ty m e a s u r e m e n t s and ana lyses ; the center zone, however , remained

r

intact ( s e e Fig. 18). two outer zones e

to the re la t ive porosi ty of the th ree zones ( s e e Table XVII). zone was consis tent ly the least dense (8.7 g /cm3 ave rage ) and the middle zone the m o s t dense (15.3 g/cm3 average) . (11.8 g /cm3 ave rage ) was between the cen te r - and middle-zone va lues , The pos t i r rad ia t ion density of the middle zone f r o m ND43 was g r e a t e r than the p re i r r ad ia t ion density of the fuel. middle zone a r e re la ted to the low porosi ty and also to a d e c r e a s e i n z i r - conium content.

Apparently the center zone was not as b r i t t l e as the Density measu remen t s confirmed v isua l observat ions as

The center

. The outer-zone densi ty

The high densi ty values for the

It is hypothesized that the annular zones and the i r attendant c h a r - ac t e r i s t i c s developed because of phase changes that occu r red i n the fuel alloys when heated and cooled in the r eac to r . The uni r rad ia ted alloys undergo phase changes at t empera tu res of -595 and -655"C, which aFe within the range achieved by the fuel during i r rad ia t ion . a r i e s can , t he re fo re , be considered as i so the rms of the fuel. Below 595"C, the phases p re sen t a r e a ( U ) , G(UZr,), and c(U,Pu) , although s o m e y ( U ) may b e retained on cooling. (bcc g a m m a ) exists. phase changes at 595 and 655°C a r e complex. tem four-phase react ion planes a r e p re sen t at each of these t r ans fo rma t ion t empera tu res . y t p * a t c . alloy and d isappears at only sl ightly m o r e elevated t empera tu res to give s ingle-phase gamma.

The zone bound-

Above -655°C essent ia l ly only a single phase This phase is a solid solution of U, P u , and Z r . The

In the U - P u - Z r t e r n a r y s y s -

At 595°C the reac t ion is a t y =+ 6; at 655°C the reac t ion i s The quantity of be ta phase ( U ) is small in the U-15Pu-1OZr

In relat ing the fuel s t r u c t u r e s to the calculated t e m p e r a t u r e d i s t r i - butions ( s e e F igs . 4 and 5) , the uncertainty fac tors in the calculations m u s t be recognizedOz3 ( f i s s ion rates), the coolant flow rate, and the thermal Conductivity of the fuel. The uncertainty in the calculated max imum t e m p e r a t u r e at the fuel cen ter is +53"C at the beginning of i r rad ia t ion , and would be h igher at t he end owing to the uncertainty i n the reduced t h e r m a l conductivity of the fuel, The uncertainty in the max imum inside cladding t e m p e r a t u r e is 537°C. Without the application of uncertainty f ac to r s to the t empera tu re prof i les , the calculations indicate that the bot tom of the fuel e lement would not be above 595°C and, t he re fo re , would not undergo a phase change. with the observat ions of e lement ND35, as shown in Fig. 9. of this e lement t h r e e zones w e r e p re sen t , indicating a fuel su r f ace t e m p e r a - t u r e of less than 595°C. a l so a g r e e s with the prof i le of Fig. 4. At the top of the fuel column, how-

corresponding sect ions in Fig. 9 show th ree zones; thus, the t e m p e r a t u r e of the fuel su r f ace was lower than calculated. This difference can be ac- counted fo r by the 37" uncertainty i n the cladding t empera tu re .

These uncertaint ies a r i s e mainly in the neutron flux

This a g r e e s At the midplane

- Within the calculational e r r o r , this t e m p e r a t u r e

The 0 e v e r , only two zones would be expected f r o m the calculated profile,

53

The development of porosi ty in ' t he fuel reduces the effective the r - mal conductivity, and, t he re fo re , i n c r e a s e s the extent of the high- t empera tu re zone. However, s o m e sodium probably penetrated into the fuel by the same paths by which the f i ss ion gas escaped. The filling of voids by sodium would i n c r e a s e the overa l l t h e r m a l conductivity of the fuel and d e c r e a s e the t empera tu re . Measurements of the net displace- men t of bond sodium into the plenum of the element w e r e inconclusive in determining the deg ree of penetrat ion of sodium into the fuel. an evaluation of the effect of sodium penetrat ion into voids on the t empera - t u r e profile has not been possible.

Consequently,

The lowering of fuel su r f ace t empera tu re by (1) lower f i ss ion r a t e , ( 2 ) fuel swelling to the cladding, and ( 3 ) heat lo s s through the ends of the element , in combination with the i n c r e a s e in center t e m p e r a t u r e brought about by the development of porosi ty , is believed to be adequate to explain the axial shape of the annular zones in terms of the phases p re sen t at r e - ac tor operat ing t empera tu re .

The development of the physical differences between the zones is not c l ea r . Under s teady-s ta te operating conditions the movement of po res and gas bubbles out of the middle zone to both the center and the outer zones would not be expected. Only movement in one d i rec t ion , e i ther up o r down the t e m p e r a t u r e grad ien t , should occur under s teady-s ta te conditions. The t e m p e r a t u r e gradient is init ially about 500"C/cm, but with the development of porosi ty it r i s e s to about 90O"C/cm. center of the fuel will always b e hot ter than the sur face . keep the gradient i n the s a m e direct ion even during r e a c t o r shutdown, The movement of poros i ty f r o m the middle zone to e i ther o r both of the center and outer zones is probably caused by the phase t ransformat ions that occur during the t e m p e r a t u r e cycling of the r eac to r . The phase- t ransformat ion boundaries (new g r a i n boundar ies ) would sweep outward during r eac to r s t a r tup and inward on shutdown, The g ra in boundaries would act as s inks for both po res and gas bubbles , and bubbles would tend to move with the g r a i n boundaries . The total number of r eac to r cycles was not l a r g e , however . which 57 w e r e shutdowns f r o m full power (45 MW). A l a r g e number of the power d e c r e a s e s w e r e at a s c r a m rate of -45 MW/sec , whereas power in- c r e a s e s w e r e genera l ly s low at -0.68 MW/min. volved in the movement of porosi ty are (1) the kinetics of phase t ransformat ion , ( 2 ) changes in the phase- t ransformat ion t empera tu res with compositional changes, and (3) changes i n the fuel t empera tu res as influenced by the d i s - tr ibution of porosi ty .

Once i r rad ia t ion h a s s t a r t ed , the Decay heat will

A total of 69 power d e c r e a s e s of 20 MW o r g r e a t e r occu r red , of

Other fac tors l ikely in-

.

5 5

E. Fis s ion-product Distribution and Composition Change Q *

The be ta -gamma autoradiographs of the polished sect ions of the fuel e lements (see Fig . 9 ) showed a g r e a t d i s s imi l a r i t y in the radioactivity of the three m a j o r annular zones. F iss ion- product migra t ion c l ea r ly occur red in the r ad ia l direct ion.

The middle zone was the l e a s t active.

The migra t ion of z i rconium and technet ium was indicated by the analyses ( s e e Fig. 19) of the separa ted annular zones. s ion product as well as an or iginal consti tuent of the fuel. nu lar sect ions f r o m ND30 was a l so analyzed fo r neodymium, with the r e su l t s given under Sect. VI. pos t i r r ad ia t ion Examination. i u m was similar to that of both technet ium and zirconium. had an appreciably s m a l l e r amount of f iss ion products than ei ther the center o r the outer zones. The da ta indicate that migra t ion proceeded about equally i n e i ther direct ion f r o m the middle zone. ,This apparent l ack of d i sc r imina - t ion a rgues against a s teady-s ta te phenomenon, s ince a s teady s t a t e at r e - ac to r operating power would be expected to promote a tom movement in only one direct ion. does not appear to be an adequate explanation fo r the dis t r ibut ion of f i ss ion products and z i rconium under present concepts of a tomic movement . number of r e a c t o r cycles was relat ively few and the number of a toms moved l a rge .

Z i rconium is a f is- One set of an-

The behavior of neodym- The middle zone

Movement of phase boundaries that a r e due to r e a c t o r cycling

The

A plausible explanation for the behavior of the z i rconium and solid

U

0 ORIG. COMP U-I57oPu-97~zI

u-147. Pu- 12 7.ZI 0 N D X ZONE

$50 A OUTER ZONE

CENTER ZONE

70

Fig. 27. Compositions of the Annular Zones in the U-Pu-Zr System

f iss ion products in the middle zone cannot be given at this t ime .

The compositions of the annular zones in a tom percent of the pr incipal consti tuents (uran ium, plutonium, and z i rconium) a re shown on a t e r n a r y d i a g r a m (see Fig. 2 7 ) . The locations of the compositions of the t h r e e zones with r e spec t to the or iginal compositions indicate s ig- nificant changes. The plutonium composition seems to have changed the least. In the middle zone the uranium concentration has inc reased and the zirconium concentration de- c reased . In the center zone the u r a - nium concentration h a s dec reased

and the z i rconium concentrat ion increased . The t rend in the outer zone is in the same di rec t ion as the center zone, but is not as well defined. A mass balance with r e spec t to the composition of the material i n the annular zones was not possible because of the lo s s of small f ragments during separa t ion of the annular sect ions. be explained on the bas i s of the migra t ion of only the z i rconium a toms ,

Never the less , the change in compositions might

56

af te r accounting for potential inaccurac ies in the data. or iginal composition was 65 a /o U , 13 a /o P u , and 22 a /o Z r , removal of 19 zirconium a toms out of eve ry hundred fuel a toms changes the composi- t ion to 80 a /o U, 16 a /o P u , and 4 a / o Z r , which compares quite favorably with the middle-zone composition. a r e added instead of subt rac ted , the composition becomes 55 a /o U , 11 a / o P u , and 34 a /o Z r , which is in the range of outer-zone compositions. How- eve r , the or iginal ra t io of a toms of uran ium to plutonium should r e - ma in at about 5:1, i f th is concept is c o r r e c t . indicate ra t ios that v a r y f r o m about 3.1 to 8:1, the migra t ion p rocess is apparent ly m o r e complex than this s imple explanation,

F o r example, i f the

S imi la r ly , i f the 19 z i rconium atoms f

Since the analytical da ta

A change in the ra t io of uran ium to plutonium, as noted above, should not have a significant effect on the Doppler coefficient of react ivi ty in a fast-power r eac to r with U - P u - Z r fuel (uran ium not enr iched) . e r a l , the magnitude of the negative react ivi ty change due to the Doppler effect is inc reased when the ra t io of 238U to 239Pu is inc reased and when the effective t empera tu re of the fuel is increased . In the center zone of the fuel, the d e c r e a s e of the ra t io from 5 : l to 4 : l o r 3 : l would d e c r e a s e the magnitude of the negative Doppler response. However, the i n c r e a s e i n the plutonium content ( see Fig. 19) i n the same region would resul t i n i nc reased hea t genera t ion and higher t e m p e r a t u r e s , and, hence , a g r e a t e r negative react ivi ty change because of the Doppler effect. oppose each other in the center zone. r a t io of uran ium to plutonium i n c r e a s e s to 8:1, but the plutonium concen- t ra t ion d e c r e a s e s , Again the opposing changes of the Doppler effect tend to nullify each o ther . The ra t io of uran ium to plutonium i n the outer zone d e c r e a s e s to 3.1 o r 4.1 but the plutonium concentration is relat ively un- changed, so s o m e tendency toward a l e s s negative Doppler r e sponse might exist in this zone. However , the development of porosi ty will d e c r e a s e the the rma l conductivity and i n c r e a s e the overa l l fuel t empera tu re , thus pro- ducing a g r e a t e r negative Doppler response .

In gen-

T h e s e two t rends S imi la r ly , i n the middle zone the

The compositional changes would r e su l t in changes i n the phase- t ransformat ion and solidus t empera tu re of the m a t e r i a l i n each zone. These changes would affect the meta l lurg ica l cha rac t e r i s t i c s of the fuel, The melt ing t empera tu res of the m a t e r i a l s i n the zones have been es t i - mated f r o m t e r n a r y solidus i so the rms3 as follows: middle zone, -1050°C; outer zone, -1200°C. The solidus t empera tu re for the or iginal fuel compositions was -1 155°C. Est imat ion of the so l id-s ta te phase- t ransformat ion t empera tu res has not been at tempted because of the lack of exper imenta l da t a for this t e r n a r y sys t em. zone apparent ly contains nickel f r o m the cladding m a t e r i a l s s ince one analysis of the outer zone of ND35 (Hastelloy-X) indicated 4,4 w/o nickel, Nickel f o r m s eutect ics with both uran ium (mp 740°C) and plutonium (mp 465"C), and would be expected to lower the solidus t empera tu re of the outer zone, No evidence of melting has been observed , however .

cen t r a l zone, -1250°C;

In addition, the outer

.

57

Q F. BurQ-up Analyses

Burnup samples w e r e taken f r o m top, middle , and bot tom locations of the fuel pins of ND32, 43, 30, and 35. t i u m ana lyses and expres sed as a tom percent of the heavy e l emen t s a r e given i n Table XVIII. Although technet ium migra t ed radial ly i n the fuel c r o s s sec t ions , no evidence ex is t s f o r axial migrat ion. prof i les f r o m the r e s u l t s in Table XVIII a g r e e well with the g a m m a p ro - f i les i n Fig. 6 and with the EBR-I1 flux prof i les . tend to be lower than the calculated burnups of the equivalent middle s e c - t ion by 0.2-0.4 a / o fo r ND32, 43, 30, and 35.

Burnup r e su l t s based on techne-

The axial burnup

The analyzed burnups

The dis t r ibut ion of total f i s s ions shown by these burnup r e su l t s in- d ica tes that the calculated axial t e m p e r a t u r e s ( s e e Fig. 4 ) , which w e r e based on cosine s y m m e t r y f o r the f i ss ions , should be shifted to re f lec t the nonsymmetr ica l distribution. that the f i s s ion rates w e r e lower than those used i n the calculat ion by 5-1070 and tha t the cen t r a l fuel t e m p e r a t u r e s might be less than those calculated by as much as 40°C. a re within the var ia t ion allowed by uncertainty fac tors .

The lower analyzed burnups a l so imply

These burnup and t e m p e r a t u r e r e s u l t s

G. Cladding and Fuel - Cladding In te rac t ion

In t e r f ace regions between fuel and cladding (Type 304L s t a in l e s s s t e e l with U-15Pu-9Zr , Type 316 s t a in l e s s s t ee l with U - 1 4 P u - l 2 Z r , and Hastel loy-X with U-15Pu-9Zr ) a r e shown in Fig. 20. React ion zones about 5 mils deep a r e evident i n both Type 304L s t a in l e s s steel and Hastelloy-X, This is equivalent to about 2670 of the wall thickness for Type 304L s ta in- less s t e e l and 3370 f o r Hastelloy-X. The reac t ion zone i n Type 316 s ta in- less s t e e l is only about 1 / 5 of that i n the o ther two claddings analyzed, and is equivalent to about 7% of the wall thickness . In the course of examina- t ion, cons iderable var ia t ion was noted i n the depth of penetrat ion i n adjacent a r e a s of the cladding. Degree of contact, t i m e in contact , and t e m p e r a t u r e are the impor tan t f a c t o r s , a p a r t f r o m the na ture of the m a t e r i a l s , in d e t e r - mining depth of penetrat ion of the reac t ion zones. The h igher t e m p e r a t u r e s toward the top of the fuel column promoted g r e a t e r reac t ion depth ( s e e F ig . 9) . t o s e p a r a t e f r o m the unaffected metal. o r 316 s t a in l e s s steel.

The reac t ion zone i n Hastel loy-X [see F ig . 2O(c)] has a tendency This was not observed i n Types 304L

F i g u r e 21(a) shows an apparent shear c r a c k f r o m the outer su r f ace The ave rage

It seems unlikely that this

T h e r e was no evidence tha t f iss ion gas had escaped through this

to the boundary of the r eac t ion zone in Hastel loy-X cladding. d i a m e t r a l s t r a i n i n this e lement was only 0.170. small s t r a i n would r e s u l t in c racking , but l a r g e r loca l s t r a i n s cannot be ru led out. defect i n the cladding.

i \

58

F igure 21(b) shows incipient t e a r s on the inne r su r face of Type 316 s ta in less s t ee l cladding. r icat ion defects o r i r radiat ion-induced defects . change of this fuel e lement was t 0 . 3 mil (-002y0 s t r a in ) . loca l d i ame t ra l i n c r e a s e s (but not g r e a t e r than 1 mil) w e r e observed , the deg ree of s t r a i n has not been re la ted to the observed inner su r face defects. Of the th ree materials] only the Type 304L s ta in less steel was m o r e mag- netic af ter i r rad ia t ion . o r possibly mar t ens i t e , i n the i r r ad ia t ed Type 304L s ta in less steel.

These t e a r s have not been identified as ei ther fab- The average d i ame t ra l

Although l a r g e r

This indicates an i n c r e a s e in the quantity of f e r r i t e ,

The ha rdness of the i r r ad ia t ed Type 304L s ta in less s t ee l had in- c r e a s e d to the equivalent of that f r o m 20- to 3070-cold-worked Type S04L s ta in less s t ee l (215-310 DPH). P resumab ly t h e r e was a l so a cor respond- ing i n c r e a s e in the s t rength o f t h e s ta in less s tee l ,

The e lec t ron-microprobe analyses of the i r r ad ia t ed cladding ma te - rials indicated that l o s s of nickel f r o m the reac t ion zone was common to all th ree m a t e r i a l s . Changes in other cladding consti tuents were minor . T h e r e w e r e , however , apparent differences i n the extent to which the r e a c - t ion l a y e r s had picked up f i ss ion products and f iss ionable a toms.

VIII, CONCLUSIONS

Sixteen fuel e lements with U - 15Pu-9Zr and U - 14Pu- 12Zr alloy fuels have been i r r ad ia t ed to 4.5 a / o burnup i n EBR-I1 without fa i lure . was sodium bonded to Types 304L and 316 Stainless s t ee l , Hastelloy-X, and Hastelloy-X-280 cladding. Maximum cladding t empera tu res during i r r a d i a - t ion ranged f r o m 605 to 650°C.

The fuel

Changes in the d i a m e t e r s lengths, and volumes of the fuel e? ements indicated the p re sence of small s t r a i n s in the claddings. c i rcumferent ia l s t r a i n s w e r e 0.5yoJ,, the max imum axial s t r a i n s 0.13y0, The s t r a ins a re at t r ibuted to swelling of the cladding and to t h e r m a l rateheting.

The max imum

. . .

Fue l geometry changed both by axial movement and by rad ia l swel l - The axial movement consis ted of lifting f r o m the bottom plug, s e p a r a - ing,

t ion of the fuel column, and lengthening (usual c a s e ) o r shortening of the fuel. ential t he rma l expansion of the fuel during t h e r m a l cycling. did not appear to contribute to d i ame t ra l i n c r e a s e s in the cladding, although the fuel completely filled the sodium annulus and was i n contact with the cladding. densi ty indicates that the cladding r e s t r a ined the swelling to s o m e extent. Swelling is a t t r ibuted to solid f iss ion products , to f iss ion-gas bubbles, and to mechanical tear ing of the fuel, the l a t t e r evident only in thecoo le r regions

The geometr ica l changes are attr ibuted to both swelling and d i f fe r - Fue l swelling

The greater total fuel swelling in the e lements with lower s m e a r

.

ACKNOWLEDGMENTS

The authors expres s their appreciat ion to H. F. Jel inek and A. G. Hins fo r fabr icat ing the fuel e lements for this i r r ad ia t ion experiment . a l so appreciat ive of the a s s i s t ance of F. J. Karasek in expanding the s t a in - l e s s s t e e l tubing. The personnel at EBR-I1 are thanked for t he i r efforts i n i r r ad ia t ing the e lements . ble a s s i s t ance i n quali ty control of the e lements before i r rad ia t ion . I. R. K r a s k a was v e r y helpful i n taking neutron radiographs. D. E. Walker and S. M a t r a s contributed the metal lographic examination of the etched i r ra- diated and uni r rad ia ted spec imens of cladding. The e lec t ron mic roprobe ex-

W e extend o u r thanks to all of t hese and to the hot-cel l ope ra to r s and m e m b e r s of our own group who contributed significantly to this experiment . Our thanks a r e a l so extended to M. A. Wahlgren, F. R . Lawless , D. J . Rokop, H. B. Evans , K. L. Ault, F. R. Kelly, and E. A. Huff of the Chemis t ry Division for the many ana lyses made .

W e a r e

C. J. Renken and N. P. Lapinski provided invalua-

a

e aminations w e r e done by D. R. O'Boyle and J. E. Sanecki.

59

The development of the annular zones in the fuel is principally t e m - pe ra tu re dependent, and appears to be re la ted to phase t ransformat ions in the fuel alloy. Recur ren t phase t ransformat ions caused by r e a c t o r cycling a r e suggested as an explanation for the physical differences between the zones. The porosi ty produced by f i ss ion gas i n the fuel would lower the t h e r m a l conductivity of the fuel, but penetration of bond sodium into the voids would tend to compensate fo r the effect of the voids on fuel t e m p e r a - t u r e s during operation.

Changes i n fuel composition and migra t ion of f iss ion products w e r e assoc ia ted with the formation of the annular zones in the fuel. The mecha- nisms involved in the movement of consti tuents a r e not known at this time. The movement of z i rconium to the center of the element would raise the sol idus t e m p e r a t u r e of the alloy in this region, where the t e m p e r a t u r e is highest during operat ion. observed should have l i t t le influence on the Doppler effect.

The migrat ion of uran ium and plutonium that was '

React ions between the U - P u - Z r fuel and the cladding materials w e r e least with Type 316 s t a in l e s s s tee l , i n which the max imum thickness of the r eac t ion zone was 770 of the cladding thickness . ove r a t empera tu re range of 530-650°C. The z i rconium content of the fuel apparent ly did not affect the fuel- cladding react ion.

The reac t ions occur red

The lack of changes in the cladding dimensions that can be d i rec t ly a t t r ibuted to fuel swelling suggests that burnups higher than 4.5 a /o can be achieved with a low s m e a r densi ty , sodium-bonded me ta l fuel. In - r eac to r composition changes fuel-cladding reac t ions could, however , be l imiting fac tors i n the U - P u - Z r s y s t e m .

differential t he rma l expansions in the al loys, and

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