ML20041D910

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Interpretation of Damage to Flip Fuel During Operation of Nuclear Science Ctr Reactor at Tx A&M Univ.
ML20041D910
Person / Time
Site: 05000128
Issue date: 12/31/1981
From: Randall J, Simnad M, West G
GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
To:
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ML20041D909 List:
References
GA-A16613, NUDOCS 8203090527
Download: ML20041D910 (55)
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Text

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  1. r 3 u GA A16613 l l

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INTERPRETATION OF DAMAGE TO THE FLIP FUEL l

DURING OPERATION OF THE NUCLEAR SCIENCE i i CENTER REACTOR AT l l

1 TEXAS A&M UNIVERSITY l P

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M. T. SIMNAD, G. B. WEST, J. D. RANDALL, l

W. J. RICHARDS, and D. STAHL l

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GA A16613 i

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l INTERPRETATION OF DAMAGE TO THE FLIP FUEL i DURING OPERATION OF THE NUCLEAR SCIENCE Y CENTER REACTOR AT TEXAS A&M UNIVERSITY s

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by M. T. SIMNAD, G. B. WEST, J. D. RANDALL,*

" W. J. RICHARDS,** and D. STAHL***

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[ **Argonne National Laboratory, Idaho Facilities i;

      • Present Address: Battelle Memorial Institute, Columbus, Ohio 1

GENERAL ATOMIC PROJECT 4314 DECEMBER 1981 r- - -

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e CONTENTS

1. INTRODUCTION ................ . . . . . . . . . . . 1 1.1. History of Core With FLIP Fuel .. . . . . . . . . . . . . . I 1.2. Observed Damage in the Fuel Elements . . .. . . . . .. . . 2 1.3. Postulated Causes of the Damage . . . . . . . . . . . . . . . 8 1.4. Preliminary' Investigations .. . . . . . . . . . . . . . . . 8
2. EXPERIMENTAL INVESTICATIONS . . . . . . . . . . . . . . . . . . . . 11 2.1. Diametral and Bow Measurements .. . . . . . . . . . . . . . 11

~

2.2. ticutron Radiographs . ............... . . . . . 16 2.3. Metallography and Alpha Autoradiography . . . . . . . . .. . 21 2.4. Gross and Isotopic Gamma Scans . .. . . . . . . . . . . . . 36

3. COMPARATIVE OPERATIONS AND RESULTS AT OTHER FACILITIES . . . . . . 38 3.1. University of Wisconsin . . . .. .. . . . . . . . . . . . . 38 3.2. Washington State University . . . . . . . . . . . . . . . . . 38 3.3. General Atomic Company ...... . . . . . . . . . . . . . 39 3.4. Annular Core Pulse Reactor Fuel Test Program . . . . . . . . 40
4. DISCUSSION OF POSSIBLE MECHANISM .. . . . . . . . . . . . . . . . 43
5. RECOMMENDED OPERATIONAL PROCEDURES .. . . . . . . . . . . . . . . 48
6. FUEL TEMPERATURE SAFETY LIMIT . . .. . . . . . . . . . . . . . . . 50
7. OPERATIONS AT TEXAS A&M SINCE SEPTEMBER 1976 . . . . . . . . . . . 51
8. REFERENCES .. .. ............. . . . . . . . . . . . 52 FIGURES
1. Core III-A . . . . .......... . . . . . . . . . . . . . . 4
2. FLIP fuel element spacing in the NSCR core near transient rod . . 5
3. Water gap and fuel spacing of damaged FLIP elements adjacent to transient rod . ............ . . . . . . . . . . . . 6
4. Profile of maximum damaged element no. 7459 in direction of maximum bowing . . ............. . . . . . . . . . . . 7 e

lii

FIGURES (Continued) .

5. Diametral variations of most damaged fuel at reference O' .... 14 ,
6. Diametral variations of most damaged fuel at reference 90* . ... 15
7. Measured bowing of TRIGA element . . . . . ............ 17
8. Neutron radiograph of damaged TRIGA element showing locations at which samples were taken ..... . ............. 18 .
9. Neutron radiograph of third most damaged element ......... 19
10. Neutron radiograph of second most damaged element ...... .. 20
11. Periscope macro of Sample 1 with alpha autoradiograph below ... 22
12. Macro and alpha radiograph of Sample 2 . . . . . . ...... 23
13. Sample 3 macro shows f ragmentation typical of distressed at ias . . 24

'14. Sample 4 appears the least damaged of those samples that show hydrogen depletion . . .. . . . . . . . ............. 25

15. Control Sample (5) is not badly cracked and central rod appears undamaged in this macrograph . . . . . . . ........ 26
16. Microstructure of central zirconium rod from Sample 1 ...... 27 ,
17. Fuel from the hydrogen-depleted area of Sample 1 is shown at left. Structure is believed to contain significant quantities of low hydrogen alpha phase. Fuel at right is delta phase ZrH 1 .6 . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
18. Stainless steel clad from Sample 1 . . . ............. 28
19. Segment of Sample 2 showing porous region ............ 29
20. Microstructure of zirconium rod in Sample 2 ........... 30
21. Typical fuel structure in areas away from porous region . .

of Sample 2 . .. . . . . . . . . . . . ............. 30

22. Sample 2 fuel from porous area . . . . . ............. 31
23. Microstructure of zirconium rod from Sample 3 .......... 32
24. Two fuel structures were found in Sample 3 similar to Sample 1 . . 32
25. High hydrogen microstructure is present in the central zirc alum rod of Sample 4 . .... . . ... .......... 33
26. Sample 4 fuel is shown in this micrograph ............ 33
27. Sample 5, central zirconium rod . . . . ............. 34
28. Typical fuel from Control Sample (5) . . . . . .......... 34
29. Stainless clad from control Sample (5) appears undamaged . . ... 35
30. Gross gamma scan of most damaged fuel element .......... 37
31. Qualitative graphic description of possible changes in hydrogen distribution leading to fuel damage . .......... 46 iv e

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i TABLES f .

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1. Pulsing history of core III-A . . . . . . . . . . . . . . . . . . . 3 I

2.. Significant measured and calculated temperatures in core III-A . . . . . . . . .......... . . . . . . . . . 9 l 3. TRIGA fuel diameter measurements at reference 0* . . . . . . . . . 12 J

! 4. TRIGA fuel diameter measurements at reference 90* . . . . . . . . . 13 l S. Summary comparison of operational history . . . . . . . . . . . . . 41 i

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1. INTRODUCTION In a letter report (Ref.1) dated November 1,1976, Professor J. D.

Randall of Texas A&M University presented the history of core operation, the discovery of the damaged FLIP fuel, the description of this fuel, and the possible causes contributing to the damage of the fuel. The results of the preliminary investigations designed to interpret the damage were also pre-sented. The following is a summary of the salient points made in that report.

1.1. HISTORY OF CORE WITH FLIP FUEL From June 1973 to September 27, 1976, the Nuclear Science Center Reactor (NSCR) operated with a mixed TRIGA core (Core III) consisting of 35 new FLIP elements and 63 partially depleted standard TRIGA elements. The core operated 287 MW-days in the steady-state mode and underwent 725 pulses.

The pulse reactivity insertion was limited to S2.00 until July 1975, when the technical specifications were changed to allow pulse insertions up to

$2.70. All of the 30 pulses greater then $2.00 but less than $2.70 and 37 of the 54 total $2.70 pulses were performed in approximately two months immediately following license approval for higher pulse insertions. The peak core temperature rise was 727'c with the S2.00 pulse insertion and 883*C with the $2.70 pulse insertion. These were calculated temperatures derived from observed thermocouple temperatures located not in the hottest fuel element.

1

1.2. OBSERVED DAMAGE IN THE FUEL ELEMENTS During a loading operation cs September 27, 1976 (about 14 months after approval of pulsing up to S2.70 four " lead" elements, each in a different cluster of FLIP elements, were found to be "somewhat deformed." These ele-ments all operated in nearly the same flux. The elements in the next lower flux regions were not damaged. The four damaged elements were all posi-tioned adjacent to the transient rod throughout their operating history.

The pulsing history of Core III-A is given in Table 1. The maximum fuel deformation occurred in the west position closest to the transient rod.

This element was separated from the transient rod guide tube with a water gap of approximately 0.05 in. The damage appeared to be slightly less severe in the elements positioned south and east of the transient rod (0.204-in. and 0.075-in. water gaps), and the least damage was in the fuel element in the north position with a water gap of 0.075 in. (the only ele- ,

ment to pass the go/no-go gauge and thus not removed from service). The core configuration and details of the region under study are shown in Figs. I through 3. The calculated peak-to-average energy ratios in the four elements were very similar.

The visual inspection of the most damaged element revealed bulging in the cladding and a bow in the element around the fuel centerline. The data from a profilometer scan of the most damaged element indicated a dis-placement of approximately 0.15 in, near the midplane of the element, in the northeast direction. As shown in Fig. 4, the upper end of the element was displaced by about 0.23 in.

  • Since there were no technical specification requirements for annual fuel inspection or for inspection after newly increased operational limits, no such inspections were performed.

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TABLE 1 PULSING HISTORY OF CORE III-A(a)

Insertion Number Of (S) Pulses Worth 1.00 2 2.00 1.10 2 2.20 1.15 4 4.60 1.20 6 7.20 1.25 23 28.75 1.30 7 9.10 1.35 4 5.40 1.40 '16 22.40 1.45 3 4.35 1.50 37 57.00 1.55 1 1.55 1.60 10 16.00 1.65 2 3.30 1.70 7 11.90 1.75 32 56.00 1.80 9 16,20 1.85 3 5.55 1.90 13 24.70 1.93 1 1.93 l

1.95 4 7.80 2.00 455 910.00 2.10 2 4.20 2.15 1 2.15 l 2.20 1 2.20 2.25 4 9.00 l 2.30 1 2.30 l 2.35 1 2.35 t

2.40 2 4.80 l 2.50 16 37.50 2.60 2 5.20 j 2.70 54 145.80 l Total Pulse Worth 1,413.43

(* From Ref. 1.

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h TRANSIENT R0D WITH AIR FOLLOWER h STANDARD FUEL h SHIM SAFETY ROD WITH FUELED FOLLOWER h FLIP FUEL h REGULATING ROD WITH H2O FOLLOWER h GRAPHITE REFLECTOR h INSTRUMENTED FUEL h PNEUMATICTUBE h Sb-Be NEUTRON SOURCE Fig. 1. Core III-A (from Ref. 1)

e NOTE: MAX DAMAGED FLIP ELEMENT BOWED IN N.E. DIRECTION N

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GRID PLATE ADAPTER l S Fig. 2. FLIP fuel element spacing in the NSCR core near transient rod (from Ref. 1) 5

FUEL ELEMENT 1.411 O.D. GUIDE TUBE I.50 0.D.

LD-5 1.37 1. D.

WATER GAP .NW-0.075 11 4 511

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  • PEAK AVERAGE ELEMENT ENERGY CORE ENERGY Fig. 3. Water gap and fuel spacing of damaged FLIP elements adjacent to transient rod (from Ref. 1) 0

. . . e e c CORE III- A GRID POSITION C-5(SE)

[DIA.= 1.464 inNE ELEMENT SURFACE

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l.410 in 1.425 in 1.425 in 1.426 in 1.425 in 1.410 in GRAPHITE FUEL FUEL FUEL GRAPHITE

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O 4 il l'2 16 2O INCHES FROM BOTTOM OF ELEMENT Fig. 4. Profile of maximum damaged element no. 7459 in direction of maximum bowing (from Ref. 1)

1.3. POSTULATED CAUSES OF THE DAMAGE

  • The following factors were considered as possible contributors to the damage of the fuel elements:

e Incorrectly calculated power distributions in the core and the elements, e Water leakage into the air-filled follower of the transient rod.

  • Improper safety limit for FLIP fuel.

It was considered statistically unlikely that four defective fuel elements were received and that these were the only ones placed in the damaging regions. -

1.4. PRELIMINARY INVESTIGATIONS ,

The core peak temperatures were calculated from the thermocouple readings and from the ratio of the flux or power density in the peak core location to that in the thermocouple. The significant temperatures measured in Core III-A are listed in Table 2. The measurements of temperatures did not, however, include instrumented elements next to a voidsor a water hole.

Calculated power density ratios are used to correct the measured temperature to the peak power element and also to correct for the power distribution in a specific element next to the pulse rod. The flux distribution inside a fuel element adjacent to the pulse rod (in the location of maximum damage) was measured to verify the calculation of power peaking. The calculated values of flux peaking yielded conservative results, so that the calculated flux values appear to be higher than the measured values on the fuel sur-face. Hence, it was concluded that the calculated adiabatic temperature distributions are also conservative values.

Visual and neutron radiographic examinations of the two followers that were used during the period in question do not support the postulate that water may have leaked into the air-filled follower of the transient rod.

8 I

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TABLE 2 SICNIFICANT MEASURED AND CALCULATED TEMPERATURES IN CORE III-A(a),(b)

Pulsing

$2.00 Pulse Maximum observed thermocouple temperature 427'C Maximum core temperature (c) 764'C S2.50 Pulse Maximum observed thermocouple temperature 499*C Maximum core temperature (c/ 385'c

o. S2.70 Pulse Maximum observed thermocouple temperature 520*C Maximum core temperature (cf 920*C i Steady State (1-MW Operation)

Maximum observed thermocouple temperature 450*C Maximum core temperature (c) 575*C

(* From Ref. 1.1 i

! (b)Using a ratio of peak core power to power at the I

the rmocouple = 2.8.

(" Calculated; includes ambient temperature of 37'C.

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The question of whether an improper safety limit had been set for FLIP

  • fuel is answered in the negative. The vast bulk of experience at General Atomic and elsewhere, particularly in the comparable Washington State University TRIGA, does not suggest an improper safety limit. It was tenta-

) tively concluded that the mechanism producing the damage in the fuel ele-

, ments is related to some phenomena occurring during pulsing, and that the steady-state history of the fuel is not a factor.

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2. EXPERIMENTAL INVESTIGATIONS The investigation of the fuel damage described in this document has been a cooperative effort by Texas A&M, Argonne National Laboratory (ANL),

and General Atomic. The examinations performed are very significant and have produced indispensable information, but the budget limitations applica-ble to all concerned have predictably led to the fact that not enough infor-mation is available to answer all the questions raised during the investiga-tion. However, the necessary information for a qualitative examination and evaluation has been obtained.

The tests and measurements that have been carried out on the fuel elements at Texas A&M and at ANL-West since September 1976 have included metallography, alpha autoradiography, neutron radiography of the damaged fuel ele.nents and the transient rod air-filled follower, diametral and bow measurements of these elements, and determinations of the axial flux distri-butions by means of gross gamma scans and isotopic gamma scans. The two diameter measurements were at 90* from each other and each at approximately 45* from the direction of maximum bow.

2.1. DIAMETRAL AND BOW MEASUREMENTS The diametral variations of the most damaged fuel element are shown in Tables 3 and 4 and in Figs. 5 and 6 at 90* apart. Figure 5 exhibits a sharp increase in diameter at 9 in, from the top shoulder of the element. The diametral increase at this point is approximately 0.08 in. This peak drops rapidly to 0.02 in, at 10 in from the top shoulder. There is then a shal-lower (about 0.04 in.) and wider peak in the region between 11 and 13 in.

This peak diminishes sharply to a plateau of about 0.02 in. in the region between 13 and 18 in. There is then a gradual decrease in diameter up to 20 in. from the top shoulder to reach the original diameter.

11

TABLE 3 .

TRICA l'JEL DIAMETER MEASUREMENTS AT REFERENCE 0*(a)

Inches From Diameter Top Shoulder E.F. Pin Forward 1 1.41205 1.5 1.41205 2 1.41242 2.5 1.41235 3 1.41321 3.5 1.41450 4 1.41291 4.5 1.41606 5 1.41600 5.5 1.41193 6 1.41850 6.5 1.42150 7 1.42135 7.5 1.42702 8 1.43150

  • 8.5 1.44344 8.75 1.46272 9 1.49217 9.5 1.43195 .

10 1.43075 10.5 1.43358 11 1.45000 11.5 1.44712 12 1.45073 12.5 1.43834 13 1.43404 13.5 1.43225 14 1.43150 14.5 1.43149 15 1.43063 15.5 1.43005 16 1.42916 16.5 1.43000 17 1.42855 17.5 1.42872 18 1.42546 18.5 1.42283 19 1.42012 19.5 1.41708 20 1.41478 20.5 1.41480 21 1.41455 21.5 1.41323 22 1.41308 22.5 1.41312 23 1.41330 (a)

From Ref. 2.

4 12

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TABLE 4 TRIC4 FUEL DIAMETER MEASUREMENTS AT REFERENCE 90*(a)

Inches From Diameter Top Shoulder E.F. Pin Up ,

1 1.41194 1.5 1.41215 2 1.41060 2.5 1.41075 3 1.41145 3.5 1.41150 4 1.41125 4.5 1.41090 5 1.41000 5.5 1.410302 6 1.41000 6.5 1.41?13 7 1.41260 7.5 1.41225

- 8 1.41151 8.5 1.41130 8.75 1.41081 9 1.41490 9.5 1.42668 10 1.44605 10.5 1.44433 11 1.43375 11.5 1.42926 12 1.42285 12.5 1.46073 13 1.43510 13.5 1.43070 14 1.43836 14.5 1.43152 15 1.43326 15.5 1.43865 16 1.44500 16.5 1.43649 l 17 1.41776 17.5 1.41628 18 1.41445 l 18.5 1.41290 l 19 1.41290 l 19.5 1.41158

! 20 1.41073 20.5 1.40850 21 1.40901 21.5 1.40951 22 1.41028 22.5 1.41070 23 1.40948

(a)

From Ref. 2.

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(approximately 45* from direction of maximum bow)

( f rom itef . 2)

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(approximately 45* from direction of maximum bow)

(from Ref. 2)

}

The diametral changes shown in Fig. 6 are at 90' to those of Fig. 5.

  • Here the diameter increases sharply by 0.035 in, at 9 in. from the top shoulder, drops continuously by 0.025 in, to 12 in from the top shoulder, then peaks sharply by 0.05 in, at 12.5 in., peaks modestly at 14.5 and 16.5 in. from the top shoulder, and finally drops sharply first to 16 in. and then gradually to the original diameter.

The measured bowing of the fuel element is illustrated in Fig. 7. The maximum bow is approximately 0.11 in, at a distance of 15 in. from the bottom, i.e., about halfway up the element.

2.2. NEUTRON RADIOGRAPHS The distressed regions in the fuel appear as light spots on the neutron radiograph prints, indicating a depletion of hydrogen at these locations.

Figure 8 is a print of the neutron radiograph of the most damaged element,

  • and it indicates the locations at which samples were taken. The light areas are depleted in hydrogen, but the degree of depletion is unknown. More recent radiographs are shown in Figs. 9 and 10, depicting the third and second most damaged elements, respectively. The third most damaged element shows hydrogen depletion in the top segment of the fuel, in a fairly small region. The central axial 0.25-in.-diam region, where the zirconium rods are present, also reflects a reduced hydrogen content (difficult to see on the reproductions). Evidently, the maximum temperature was too low to drive hydrogen from the fuel to the unhydrided zirconium rods in the central axial hole. On the other hand, the neutron radiographs of the second most damaged element show significant hydrogen movement into the central rirconium rods, indicated by apparently equal hydriding of the fuel and the central zircon-ium axial rod (again difficult to see on the reproductions). The most damaged element exhibits large, significant depletions of hydrogen in all three segments, particularly in the central region (Fig. 8).

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I t g-* -'

[' a + v. O e%&V,  ;

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~

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%. . + .

.'- / ~' *

+.

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=

+ r . . 4 1

?

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l l

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,~ .; A . .  ;

i e.

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1 -

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y _

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,. _ __ -.,.' _*1b%.

j v 4 + q .; _ . ;( ; ,. ..

+. -. ._. __ :.. . _ _ . _ - .;

1. l'1 . . ! r . . !i r .11 ' t . r . i } > li i >nd "h> ,t i l . i n t.1,'r (i i t r:t il t ( ni lt riin r td it)g r <ti>h s i it. N )

()

s

~7 x s_

e 2.3. METALLOGRAPHY AND ALPHA AUTORADIOGRAPHY s

The metallography and alpitarautoradiographics of the' most severely damaged element are shown in Figs. 11 th ough 29 (from Ref. 2), which also show metallography of the cladding. The areas examined in the fuel include four from distressed , areas andithe fifth from an essentially undamaged area at the bottom of the fuel column. The undamaged area was used as a control sample for comparative purposes. -The. distressed regions were' selected from the areas that appear as light spots 'on the neutron radiograph prints, indi-cating depletion of hydrogen at those locations. The greatest concentration of hydrogen-depleted material'was found in the area surrounding the cladding bulge. Alpha autoradiographs were made of the five metallographic samples to determine the uranium distribution in the fuel. No significant changes in uranium concentration were observed.

The metallographic examination revealed circumferential cracks near the clad on all fuel samples, as well as extensive radial and random cracking from distressed locations. The unaffected (control) region from the bottom of the fuel column had only two radial cracks and a partial circumferential crack. The condition of the zirconium rod which runs axially through the center of the fuel column is shown in several of the figures. This rod appears to be undamaged in the control sample. In the other samples the central rod -is either deformed or partly bonded to the fuel, or cracked. It is most damaged (cracked) in Sample 4, where it appears to be straining against the fuel in the least damaged of those samples that show hydrogen depletion. Swelling of the central rod by hydrogen absorption increased its diameter by 10 to 12% over that of the control sample. The microstructure of samples of the central zirconium rod shown in Figs. 20 and 25 present evidence of transformation to the epsilon hydride phase, indicating a ZrHi ,7

-material.

The distressed fuel areas exhibited both a low hydrogen alpha phase mixed with delta, and a normal delta phase (ZrHl .6). The region shown in Fig. 22 (Samp1e 2) has a highly porous area near the cladding, separated 21

  • 1 0

AYi>

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rl5 , plQ t 4.; -

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s%.

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pk .

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&; g*~~!f a#3 . iLlk.:

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f , ,.

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,- t

.a . n  %'tn.,$

o -

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'Q*. , ' ?: $ W~.-

<, . :m. ,'.. , ~. 4 yN;. ng .Mni,h,'h'N2L .3  ; .'+y9w'*R % D,'h,c* A ~ti q Y .f -

y r c. ., H y. -

g wh N ;.,';cg;

,s

.4' w'tw u- A >

.yfff2 **

l '

,g

.~

h8 Fig. 11. Periscope macro of Sample 1 with alpha autoradiograph below. It is ,

likely that the light-reflective material contains significant quantities of alpha zirconium. Zirconium rod was deformed by adjacent fuel. Location of clad weld seam is at top on all five macros. (From Ref. 2.)

22

1 i

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u e - t ,} *',; ,>; nw

&- >^s.

~

+: %

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$n /

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~- pw *w,w;y y9.

s -; mp. ws. ? ?l:'

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e g onw;jf5be K,d ,s: n ': ,,. ,, .

,. a . g --!,a;3

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m.';.y:g p

e;u~. p ~ a.Y :r:a,,r, -

<f

. , m. . _

e , , _ p.,

  • Fig. 12. Macro and alpha radiograph of Sample 2. Fuel is extensively broken up.

Material on the clad side of the circumferential crack shows different appearance. Porous area shown in Fig. 19 does not show, but location

, is indicated by arrow. (From Ref. 2.)

23

m ,-a 1

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r . . a s

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y  %, t , ;+.y n;f-4 y.

.w--c s ys .w-Fig, f3- Sample 3 m cro gyny, fragmentation tyP i cal of distregggg OEGaS. iFrom s Ref' *,-)

24

l i

, . " ry: 'd -

^$ .

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ff.(T . 4 _ . $ l vh w ' R ' ' , ' ~

~

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a s.f+.x ,(.')A.[

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+ -

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r

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4}\

..<- 4;c .

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> f,p/ . ,

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t,.

-\ q. 3., -

+;;y. ,

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4 >-

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. , . . t

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p*.3 .. or 4: . ;: a. ve :

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+ % pgey,,

sw,,

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, ' ag--.k.

s ,c

(

.--%,, , ...e. t

>A ,5 4.- ,. 4. i g + ,r

.c

.e :+ 3; w; -c. . -

.4 - ,

. ,, =' y l- 4;rv,),Qh u~

. ts

~-

.: a

.s ll ,

th..if* ,

r; w

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\

l 1

l

~

%% ?/

e

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g. .. . ' wh. .-

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e e s.-r -r p. s ..yp- '.: p p %.m>q,Q,.g o u.s.,.%g%., a,

.s# ' 4 .. ,

.-< /g N.3:4 yg

  • s '

,,,...m_.. .-

Fig. 14. Sample 4 appears the least damaged of those samples that show hydrogen depletion. Central zirconium rod is cracked more than in any of the other sections and appears to be straining against fuel. (From Ref. 2.)

25

i

'l

s. ' *,,

+,.:

  • [ ,

i

s. '

p

.t<

1*' y O. _g .. , . . .

~ , ,

Afsni3

(

~m7-v, 1-Y[f,_ ,

' N?y b*

+,i ' ;i s%

',[

I l

l 3.5X ,

I y

l g yp.u W S . g.,  :[i , /

e ' ;w Q!Qffq u - n 'i:=

,' b  ; h[k $'Ihbhh x: ., &{[- > / . sQf.?fjf';"%;6)[

, b'.: ~ & ;, ','k, , . ,

l u;*,.,.

% ,m,'w , 4.j y;.d, m _

, 6:*9;.p 3.tj-

~ ug, **

,f

>,;;y ; r ,-c; ' - w.s . ;_ . 7,
?.j;;V

. . . , , - - . . ,c -

~

_ w e~

\

y ,

Fig. 15.

Control Sample (5) is not badly cracked and central rod appears undamaged in this macrograph. (From Ref. 2.)

1.

i 26

l i

l - 1

.. s N., .

y/

~

Y .

\ U) .:.\ .cf' 3 . (.g'Q Ey$.,

y +

+ . , ,>- ;

y y(l %

s ,

A y g"\ .Q *,. ,.

\

v *

, i

\ g, ,

m f. I ' d. i

., 5. x? ' ', '.O"U hf ~

~

.U &p . yh a,. 'fV y gr

u. I .

'qTi .ng -

\L y*

.  %. a .

4' e e.a p <..

-t WJ

.. g i*.'

', ;* ,u . .hin e f ^

f .y ., " ' &_ .o

s. mu -

.c p; s.f,

- u. -

<e*tc .

I the r .+e# 4

. Pd LIGHT 800X 500X Fig. 16. Microstructure of central zirconium rod from Sample 1. (From Ref. 2.)

.w. , . - .. . . . . . - . .

, s; .. :. W, ..Y.@.N.,'.('~'- #~ '

. , . ~ . .

y:' . :q. . -

?, ;. ts';.w f ,

F. -. i..4.s  ; . . .3.4 a  :. .

.. - ~. +

'. t

..>. ,z3,._. we . :( . . -:,. .~.n

. f. . . ,... . s

.; * .?. . . - .~ ..

. . . e:. ,.

.s

.. + . >, . c ..

~ . .. , . ..

g',c.s.~.g, w .

. ..'- . .g , . .. . ;.4, 5 , *, . ' . , ,

,. . . s e ,,, '

...j. .. ,. . .

. + . ' .. . ; ; :4 .. .c ..

. n. . . .

~-

  • n .,,.g. g . s .

,,..?...,

4 s*,._... .

g  ; : ... . ,, ., .g .,

  • {

s.,,j,s ' ;-.l .. .. ~F ;

N.,.*.. . ..~. %.al.

,.~.s. *

,, , ', . . . ,.o ,

~' 3 . . . ; ' .. . .

., g .....t  ; , s/ . -; .;n. . 3 . , t'.e. ..

q' ...- . = . . , * . . .. . r. . . . s 'o . .- ..

.,,,..~;

.'*r* * * ' * * .*

s s , .*

  • ;...- .1. .

. e * . -

. e 9. : ,

, ,, ,., s -

Pd 800X B.F. 800X Fig. 17. Fuel from the hydrogen-depleted area of Sample 1 is shown at left.

Structure is believed to contain significant <[nant i tles of low hydrogen alpha phase. Fuel at right is delta phase Zril f, . (From Ref. .

2.)

27

1 l

i

,. Y$ l, '

Ilf I' k h.h,. .N h..

' '5 .,..,.,.W$Y

&,hh a h&w +?..

r\,.

  • . ' \ . gt ..

+ -

$s .

  • ' l {., . '

.'g.',.

.h% .f pch , ,?

, .('.h;~_d.^b lH ?

x .. .

,*:.7* . e. g. . . ~ . . .

c: y'f p r?-

- ' 's

, ," , , f. g., / 'i.) y,,, t l

1.*'). Q;  ?. ?-e. f,p Ig ? g,'l % '

3 k'?'h,lh)p.f... .? "! ..

hQ 5 AR ' *  ;

. fQ.( , p. . ;f. c'L. t. 'g';

e{ .

v*] i-%..v:. '.'.i w '.X -l. .A '

.c

<: n'... sf W,' y. y \ - p

'3

..-W '

{,. .

.- p :. n{ ,' c ' . ; Q.:-; J.l,3,, :r. . .' ,:e.s. :w . :

n).i.1s4.p4,4,[S;T

.

  • l M,

~o.$..

.\

e/

.s M

W,:.4,l' -.Qp.'r

(., ;

.a

.l '

O' Adu l. '. A h.&? b& $.; b5,. Qn

,$.<,.. [, Nk . 5.',$.$U?hy<:: $ } f;

Vl ,:,b h ' ,

!, L k, <a.:. , 's .. i. - - ..

l )

l,.,';.'.,t,."416V}"h}~j..

, s ' , H n. nb.  : ys. A (s.

. .w L'

f,4 t n.5:.' . '.

q%. . ' : Ly '.

3 m&ft

'. ., ts C:

nwy,. s. ,

135X 35X i .

h r ,

\f

j'; . I

,fy i" y% ..

Qn

%.mN  ; .,

"(.ti ; .. 7- ,

t ~un a3  :

4-fiQ . *

.t t lp, 9 *.:y s. y DJf:

g is$d a .11 i, J 4m. l .

.. 's

, .i

. g'ef

, J v]l f j j . . .. ,  ;

n..,w; . . . . /.e [3", ;I f

. , i 4

.Nv w. q.. , Igcp.

.w - . . .

i

. * ' ' i J j wn ' %g,,m;. '. g.. *

,) .

w' ' y ',

,.j 4 r

j.

' p,i

. j . ., 4-

'l l

' ~ J,y4 .

,j 2s'g';l'

.e

  • , t*j e , , .g ;,

f p)f

>'i a bi

. .:. t *. i , % ,

,- ^ j: ' ,

.e '

' , *,- l',_

,g ;p of.,

y . . ' .i[> '

. - e re I. >

Nl ... 3.l.. ' ? $., , :A f . .

. :1  :

%rg,' ?n j OXALIC ETCH 120X 500X Fig. 18. Stainless steel clad from Sample 1. Microstructure of welded clad has not been sensitized and appears to be undamaged. (From Ref. 2.) .

28

<r-_-,-- - , , - - - - - . . - --

1

~

l i

/

p '

l. .

. ,; 7%, ', '?n. .

jg.ys# '9 ., .

E'.,".,,...e..

. ... .'.M.g, . .

g. ,. ,* . r.e. . . ., .

y.,

g *. v T '@ .ia Y. .' . '.*. ,' .,,'),q s,'

r.. :.

- W, ...

  • ' *w.'. . < - ...-  %-

kWC j e'w/

  • s: . .& **

-~

" 5 ** - ' ' . M... "

~'

. ..! g.Y D > I.$',t,,. 7 ? D ; O. .

.,g. . t . .:.lA. - . . . . . 'Y.'\ >... ,.. ?,. ,,.,.e,

!N.f .

s . .. ,

.' s; .. q..s. . . *.. .

  • s. '.e , . ,

f,*.fe. .. -* . * . t 1 . . , - -

p ;..;,. ,y., ;

. 4 l,t , -

...? , ,, ,

r. . / .f

.; c , . .,-

,p> f..; . . L..
. ', ~ , ' *

. ,. , e ;,,s c

..  ; ,t .. ... . -s. . . .., , , ,

s . .

.*:,; f*,' ,1, ^. . . -

l(i . * ,

. " . ' ' .j - .l}..

l ,

.,P. l .

a . -- .'s. .

p- -

a .,., -

.n.,.. .,, .

'] . ! ,' . ** , ,.,

's l. , . . . :. , \ , V^ *

. =;

..- n- .

r. . .

.x

..: . > -( ;

't

, =i,-

. . . y, -

10...

t...

. e a. . '

sn.  :. . . .

w.. .

7'."

e. , ,

. _ __ , y

,.N a

L .

1 l

1 '

t I

~10X Fig. 19. Segment of Sample 2 showing porous region. Fuel between porous area and clad (clad not shown) is a uniform delta phase material; that of porous region has a two phase structure. (From Ref. 2.)

29

- - _ --__.-_-.y- - - - --,,

. .-s

.% j d,i "

. Of6 R .W . hO Yt[ ,

w w .

y

\ }\ ltl f' sl L V '

..w w lA g" g

\ so ls

%
:a.,(' .x 3 ;\

\y . QT -

kby[*h ' 4L- , giy a

y i Y ~ ;l? $

{ ,'O M. a.d .

Pd LIGHT 800X Fig. 20. Microstructure of zirconium rod in Sample 2. Chevron structure is that of epsilon phase, indicating a hydrogen content of ZrH1,7+. (From Ref. 2.)

m .. .. - , . , . . . - . . .., -.. ,

W$N..,,m. 2$j. _,,c'4'.f.?:d T.~ '.k.- : \:.t .' ,!*

pg.a.?

g~ a.

v  : .a, .n: ; .. .q ;.;.:. ;- .,..a. -;

~ . . . .

. 4. .r ,,..3.

,. .... ,y, e,

  • p;.

. . .;. ..n.>....,. .,,y.'

,.,~;>,

,.,.t-

. .. ~. : .

z. .

t v. 3 n. , ..  % *g. ' ' , . - .

  • W g * *
  • p ',%..: .We 47 ,.3.*.*,**,

.- .- 3 7.e . E. . ..*p,1. y f . . *' ,... ,. w*.- ',

a

g. +-

.  :,.~*: ,  :.a ,

t.  ; M ' b e*(.d. ,,;.. '.., ~, . . . .. s..4*.8.e. .,..

Q. .,. h .s .

f i.'

. ..:.4 .a ,

e.s (.

  • e:.s. v. .:A c

9 c. .

.s s'"

. .* t ,,

[. - .: , .*. , .* ..

". *. . * .,.A ., y ;;\ ' , .,. '..'

. , 17 ,, * .

~ .' ,.',,,A .***

j $ . ' e' . s .'.e '. ,\ d* . **' . ' .E *

*
  • n ' .-

. ,e.O'. .-

( '. h. .1.e3.~.') {.j'. . . .$ ,'.*.

,'f.,.... .

  • '~
3. . ,( ' ,',
  • f.

.i,,.,.,..

1 p ' * # .. .. ',.,

.;.. t;* .,...'*, *,. ,. .;.,*, .-.;,.  : ; .* ,,

.. . s .

s

$. .;.,.,;s, , ,, , ..

..,..3 . , , .; .. ,, .

a.... os . - - . :o .

. .;sa- -

Fig. 21. Typical fuel structure in areas away from porous region of Sample 2 (from Ref. 2) .

30

l l

. ,:S,e- s~ <v.  : :. . "; .s  :. '. .- - *- .

T ~: '.'

yi9

. s. . .

z..

. i- . . .

.~ '. K.. ;'\, 2: . 2l'y ~' '

v}- l.:. .! .': ' . l:. 4 ,. ; q. *

.r *  :- .

. ' . );;.

r .g YE.

n.$'"y',s* Q g.'pY .

g .

, :l....#.y. .;.,,C .h.,.h. :.; S. b*A&&*

.n;, O. :.. * : s... ., *

., ,s . .c*. v.f c... ".e, .

-wg'* . .  ;; .. .va , <,  ;; p n ,

7.;1 .

n .u.:

.,, .g . . ,.

.. g> .: .- ,,:,y .

. .. .y.

.~,

.r: . ..

. . .:9 . ~ .^. , . ,

s..a Q._. ,.'y:..e.:, :.'. ;;s.f x.:.: Y.,-h 1 ..- e.,, . . .

.,-[.e,.c . N ;-

. Y:- . , c- c. . m: '

~.;! - -

s -

.

  • p.

g

, . . t= -

l* e

$ * *c L /

j --

  • f. -

4.**-

. w ;' . ': -

. ..c:'58(l*.. '5s.;l:..

i . % .:'.: '

. ,U: -l$n.&. M.:  :  %'O,

,% .:..:.;.,.. . .2., . .

  • ~

R

..t %.5 *. * ' W.

.,i < E1..

. .r

.,>. '4 * .

t .

-v::.}.:.L :~ r.e

.-.r.,

t...

.,s s  ;. ' *

.r. *',cc~. D a*.. .% .

y: ; r.,(m . %,f. .<.r ; ,+',-  ; '. . .:-.; p:.. ., cm' g ,;. ..~ek,.

A.e.

- 9. :~. .w, - ..

4 :. '.e . -

.a-

v. . r :n..:. . .r..'.'o,'y.

1 ,

>. ~%5..  :.. c .w,' 1. * .

y;%. q.. t., K..

.f % .!:.r. .'..a & .. . ... !*. x's '.' + .l!.%' '*g. &c.?; si. ',

.. TW .

.'1.*h[

.. . f .%'..%.f s... .

.....'y....pr~.

L .: '.3+e .

4 . ;. ,:ek.

. ,. m e xX. 3. ....

. ,.  ;.g ,,; e-. .

9 ' \ f. .p' w. s...t. . h..y.

$ .. .  :.g. -

.f e,.4.:. .:. . e%,;'. ..-

. ?.a .f. y . , '84nam,

~$ .- .~ . . . .,,> o. .~;

. .. . .. .. a. ..

r ..:. .. w ... w > y o : ^- n .

e c .: . . . . ..- ..

,. ,. g ....

.. . . g .....n.o

. . . r . ...

..y .

a.

' ..e.~ .-'.} . .

.. . U. r:.3 i . .r5. ~ .

9.

o/

Y...~:' %. .. q . . H. h.'.y$l. e . . .9 . ,. '4 > '. . ,.' * *

.. , . , . .~<.

r.A. :q, .. ,, ; .- , .c w..,,. .e.,.

.. g .

.. . . =w.

' s e -

i, -

g . %,4 - % ..T. .:. ' % .i.. . .S

s. . *. e E' * - *. *< ..-

em%:N.es. Y - ** _ . '.

i. . s

..... . e / '

- - +- -

.:Qi~~*,&,

%.*t'*[R.* f.% $'f' N ~ jhti*t.  :' ' . f' *..b.'ey;.

b k. . . **.'e .- N.v.. &  :#- . ' ' ~: - ~~ .

t.'

. o.-< n .

. e s ,. . ~ . . . . . . . . .,

' o%. o*.f .',. ?.7 4, f .;:.c si.,..Q.,.:

> :.&. Y. . . , . r. \

. - , ' . L. '.=, &. %,i.'.'

',.J. , f W .i-. W. ,:%h*.:'.

s .

M.

  • s .e. . ..

. ' .;,,.,y,p M.'. Q'f" 4 '% *i, K? :,<* .a .** t.. . -

~

1 -

vn$' h

. cw. w. .

. ,w. .Y

. L. nb h .; . ..

-ff. k. n. . e..

800X Fig. 22. Sample 2 fuel from porous area. Two phase material is probably delta with alpha-delta cutectic on grain boundaries. Bottom right micrograph

. is of porous to sound material interface. (From Ref. 2.)

31 1

e l % f

... I\m

.s s

..,  ; ' ,b ,' ' -

f. -

4 4  ; ' $ '

's '

.. ,' c

$ f' '. g <

f -

.g -

\

l ,. ..

't. -

o

)'

\

'i s s

\s \ ' j. 'r A * '

\. , >>

\ . ..

9

. 'k ' \ . . ./

~R '

I  %[ #

s 5.

Pd LIGHT 800X Fig. 23. Microstructure of zirconium rod from Sample 3 (from Ref. 2) ,

l y o,y *c

. .e. . yyy . 1,.

. ,*.e..

. . w. m e.?..v  ; [ .' ,. e :w:,. ..W'*: 'i.;s*::' * * ;

m? .. v~ . @~

. . w ; p. , m ,.,. . .

4iM.PQ..F9 s .  %.

M S.l-

. . . 4 .?. m g ... m..

.ggw- a ..%>.

m. .

w.#. . : ..

f *5[]., N*N ul,,1*,"l,lbj,.R:', . ~n . **

.~,.

'.%'. . ?

p.s'm::J'ry*. , . .ay;.*f~.>%

.z*** b ..'y.' .- .,

.b % .f. Md

,,({.p q,.

Ms*.'f.*'

p q)J,.* j(,

7*.f,$...?.s:9^

so r.s

  • s*  ??. .

. n c .!,E.

.y % q ,?**,.*g7;; ,p?..4', .d.*

.. .'b n N * 'F@. -

F I .7.* ~

? el;*

s *f. , 4 *, %y. 1?<ae, * , p * ?p . ,g- 2 V'.. mg '1

_ 5 ,r , .

QI. ey'.6 .f;a.b("[U.'h'h

!A .'. '

b.- *'

a L' . *.'*.;&? mials.K.Mi.n .

Pd 800X B.F. 800X Fig. 24. Two fuel structures were found in Sample 3 similar to Sample 1. The i alpha (left) comprises about 20% of the area and delta the balance. *

(From Ref. 2.)

32

~

~^

, ~.' Jg' ,' ,a f

. 3 ~ '9 6

s -a i ,,

~ / W9- ^ +-

1

. . : _ dgg'4. e.4y-.}

~9r . y

' ,4 - y1 d'

.._. .'. 11 -

l

-,..,-.~,-

'j ' ~ .

7 ,

'?

%. y 5

+-

f . . [? *.

a b-3

. fi. , 2 Pd 800X Fig. 25. liigh hydrogen microstructure is present in the central zirconium rod of Sample 4. (From Ref. 2.)

~ '

  • .. :. c;  ?. e ^',k% ~.' :' ,. yy.',
2. .; .'

, p,*Y'j.<&s.

f.,,'

sh 4 .*~ . . _..'s

.. .g.=.9

'. .'E - . .

'_ . . % 4 E. l {.* -.

( .')l,-% ;;y . f ,

l

- . /. o ..' - a f*., '-  ;-

g . ,. ) ,, , t j2 . . . . -

..h. pg' ~,

, 5). > . .4 ,;

h[ , h :N4 h' j

M N

  • h ,

% . ,%, . \

i .'.<A ,;

,',,y ,ys, 5 :.*b, m: ,,?

c y

.pt.<,,

i

'9 . g em

.G{T w . '. j /,~4[ , ,,

,A' '\_ ,)..

, ig$ . +. -

2 7 ., . . , . ,

vu e i

R ,. ,

_3-

,a , g., Y.*s 4 1 .

y.,

, l., me.,- y %s , */

B.F. 800X Fig. 26. Sample 4 fuel is shown in this micrograph. Structure is primarily delta with some epsilon also present. Ilydrogen ratio is probably

. close to as-manufactured specifications. (From Ref. 2.)

i 33

,-,,_y

,- 4 -

M b - .

y

._ hy e

l ft; - * . .,

I

.' l

' s

.; ' .',- l o

Pd LIGHT 800X -

Fig. 27. Sample 5, central zirconium rod (from Ref. 2)

.. sere 4 7;$w;-

  • '.  % i.

4E g

}sf.. . ..

r . . ,%

f

. y -- c - -

3,

f. . .,

, ,'M '

d i

g ;,.

  • p. * .+

4

%. .* E,. .. } 's 9..-

,' f'

, ' O., y .'.

k- 8

*A r  ;. - , -

'~

, e.80 - )* 'u s-4

- L r L ' A ,; si u Pd 800X OXALIC ETCH 800X Fig. 28. Typical fuel from Control Sample (5). Microstructure is delta with some epsilon. (From Ref. 2.) .

34 b

, . ~ - . , , . - - - - - . . - , , --. , , , , . _ - - _ , - _ . - - , - - , - - - - - - - , - , - - - - . , - - - , - - - - - - - . - - - - - - - - . . - - - - - - - - - - ----

~

1 1

% J?.i 5 5.,8

y. -

3 . . n. '::. ,0 i .

e -j w, ,. ,

i , ,Ki"# ! -

7", ; T ' .. ,'

,5 s

g.[ ,'.

' 8-u .tg g

N

' , ,[

- -9. - 3 7 -- -

.t, ,

I l

. g , ,1 r f. . . ;. . ,

. ./' ,

, . , _ l e

r .t ., , ..

s.

l

,e, C ,. ..

t ' *' '

1 2 ,' : ' '] ;.; ; , . " , ' ~ ' .. ,.  ; '

~~ Y l ..

y -

.i .s\- _ . ,

n.

, ; e ~m  ; - < ; 1, . =. ;  ;," u

./'- i 'd ?  :, " ,c-pf.:.. ' s .

._ .. [.. .. . :

- , .'. i

."r . . . _ ;f.p. . ' -

', w : ~.-

7 Q.
* p;;.

o :

' , r ,. - .

~,

, ._. m:. .n . . . , ,

I *. A; a .4i. . *:.;- . .

l. - A .i  ?

.-: :l:' . a, . , , :-:.

. p. -

". c .:. . : ' -'

. j 35X l

., 4- ,

ff ',d .o '

  • c, y- ," ,

'i ,Ifl 2.1 f;

,.,@ ,^ * ..., ,., ..

.- .- 5- ,- as..

'-t- , .

w- '7;/m 9 y..6

. n.v, :.J a:;:m.

^.tr; 1.. -- .-

s - N .p e ., =" -

f ..gsv\ . -

+

If h, y L i .;. l. y l<$}y y.:

a.\ .,kgYr. .

'?~ %, - .

-- s <

y/.-l-- .>...

~

8 ..s.p .%y i - n *;'y ~

<  :; .{  ;. -

. . . . 3., s, 5 ./ ,,

D, 'g , .. J '/S.d..

i ,

. 4. y * .

. i ,; /9" * **

.. y 4N N.

s '. * . l . ,, ". Im' '. .:

-- .' .! . 1. , . e el t. .

120X 500X Fig. 29. Stainless clad from Control Sample (5) appears undamaged. (From Ref. 2.)

I 35 I

from the cladding by a cracked nonporous delta phase layer which extends circumferentially but nonuniformly. The two phase porous area is delta phase with alpha-delta eutectic on the grain boundaries. Approximately 20%

of the fuel in Sample 3 is primarily alpha phase, the remainder being delta.

The Control Sample (5) from the bottom of the fuel consists of a delta matrix with epsilon phase inclusions.

The stainless steel cladding is structurally sound and does not appear to have reacted chemically with the fuel rods. A small gap is present between the fuel and the cladding.

2.4. GROSS AND ISOTOPIC GAMMA SCANS The gross gamma scan of the most damaged fuel element is shown in Fig. 30. The shape of this curve reflects the burnup of the fuel as a .

function of position, using a slit width of 49.9 mils. The burnup appears as a broad distribution peaking in the center and decreasing toward the fuel ends. Small peaks occur at the ends of the fuel element; two small reduc-tions in the gamma in ensity occur at positions of fuel rod interfaces (three fuel sections, each 5 in. long, contained within the element clad).

36

W W W v Y r 5  : W W F W 4 W W GROSS SCAN OF FLIP TRIGA PIN NO. 07459 SCAN DATE 7/13/78 SLIT WIDTH 49.9 MILS SUB. ASSEMBLY TRIG A SCAN TIME 01:37:50 LIVE TIME 5.0 SEC 20 16 - -

b X

m 12 - -

z U - -

8 - -

4 - -

s 0

18.00 21.00 24.00 27.00 30.00 33.00 36.00 39.00 42.00 Z POSITION ENCODER READING X = 7.254 Y = 7.200 Fig. 30. Gross gamma scan of most damaged fuel element (from ltef. .J

m

3. COMPARATIVE OPERATIONS AND RESULTS AT OTHER FACILITIES Of the several reactor f acilities using TRIGA-FLIP fuel, only three have the similarities of core configuration and/or pulse size that place them in a category comparable with the Texas A&M facility. These are the reactor facilities of the University of Wisconsin, Washington State University, and General Atomic at San Diego.

3.1. UNIVERSITY OF WISCONSIN This facility has been operated with a core configuration very similar to that at Texas A&M in which the fuel damage occurred. While the Texas A&M f acility has control rods and a central pulse rod, Wisconsin's has control -

blades and a pulse rod. The Texas and Wisconsin facilities both operated with mixed cores containing 35 centrally located FLIP elements. Wisconsin's generated about 2 MW-days of energy and pulsed 283 pulses, with its mixed core containing 35 FLIP elements. The major dif ferences in operation between the Wisconsin core and the Texas A&M core are the relatively small energy generation at Wisconsin and, most important, the fact that the pulse size was never over $2.00. No problems have been experienced with the fuel, and the reactor is operating with a full core of FLIP elements (since June 1979). The core has operated ~24 MW-days and pulsed ~74 times.

3.2. WASHINGTON STATE UNIVERSITY The Washington State University reactor core is of particular interest because of its operation with a mixed core configuration similar to that at Texas A&M and with a pulse size up to $2.50 compared with the Texas A&M S2.70 maximum pulse. The mixed core with 35 centrally located FLIP elements was used with control blades and a pulse rod. An important difference is that Washington State's pulse rod was water followed, causing larger power 38 e

peaking factors than the air-filled aluminum follower rod used with the Texas A&M pulse rod. To date, Washington State has pulsed over 55 t0ses with insertions of $2.50. This is about the same number of pulses Texas A&M performed with insertions of S2.70 (these pulses being included in the approximately 80 pulses greater than $2.00 at Texas A&M). Temperatures measured in the instrumented elemalt at Washington State, which is in a location similar to the instrumented element at Texas A&M, compare favorably with the measured temperatures (~500*C) for a S2.50 pulse in the Texas A&M core. Calculations performed at Texas A&M indicate that a S2.50 pulse with a water-followed pulse rod would produce higher peak fuel temperatures in an adjacent FLIP element than a $2.70 pulse with an air-followed rod.

. No problems have been experienced with the fuel at Washington State.

It is significant to point out, however, that the steady-state usage of the

. Texas core is very much greater than that of the Washington State core.

Prior to the high power pulses at Texas A&M, the core had about 170 MW-days of operation, while the Washington State reactor core had approximately 8 MW'-days of operation. Washington State still operates with a mixed core, but the number of FLIP elements has been increased to 47. Typical operation is ~2 MW-days / month and only 5 to 10 routine (S2.00) pulses have been performed since the pulsing program in 1976.

3.3. GENERAL ATOMIC COMPANY TRIGA-FLIP fuel has been operated extensively at General Atomic. Only the initial tests of FLIP fuel were done in a mixed core, which contained 18 centrally located FLIP elements surrounded by standard fuel. All other operations have been performed with a full core of FLIP elements. This core was initially operated in the Mark III reactor for long-term steady-state operation at up to 2 MW. It was then transferred to the Mark F reactor for general-purpose (steady-state and pulsing) operations. The core had about 750 MW-days of burnup prior to any pulsing operations. Pulse insertions as high as $3.20 have produced measured peak . temperatures up to 550*C. A 39

P summary comparison of the operational histories of the Texas A&M core and the General Atomic core is given in Table 5.

The differences in core characteristics, such as peaking factors and temperature coefficient caused by the mixed core versus full core, are responsible for the fact that different pulse sizes and steady-state powers produce very similar operating temperatures. The primary dif ferences noted in Table 5 are that the General Atomic core had ~5 times more steady-state operation than the Texas A&M core prior to high power pulsing, whereas the Texas A&M core had 3.5 times as many high power pulses and the !Mxas A&M calculated peak pulse temperature is slightly higher.

3.4. ANNULAR CORE PULSE REACTOR FUEL TEST PROGRAM ,

The only previously observed fuel damage similar to that seen for the ,

Texas A&M fuel occurred during a high power pulse testing program at General Atomic where five or six centrally located special elements containing 8.5 wt %, 93% enriched uranium were pulsed 426 times to peak (calculated) tem-peratures of ~1150*C. This was a program for the design of the optimized pulsing fuel for the annular core pulse reactor. No significant amount of steady-state operation was conducted prior to the pulse tests. Fuel swelling occurred to the extent that the 15-mil radial gap between the fuel and the clad was filled and the clad diameter was increased about 60 to 65 mils. The swelling occurred in the region of the localized hot spot area of the fuel. Extensive cracking of the fuel was observed as well as a porous region very similar to that in the Texas A&M fuel. The porous area extended to the outer surf ace of the fuel, however, not terminating prior to an essentially undamaged, thin surf ace area as is the case in the Texas A&M fuel. No neutron radiography nor any other examinations were made of the fuel to determine loss of hydrogen from the porous region. It was concluded that the hydrogen pressure (~90 atmospheres equilibrium hydrogen pressure for a temperature of 1150*C) generated in the fuel matrix at high pulse

. a l

r 40

TABLE 5

SUMMARY

COMPARISON OF OPERATIONAL HISTORY Texas A&M General Atomic Core Mixed 35 FLIP Full core FLIP 63 Std. TRIGA 63-100 elements Maximum steady-state power 1 MW 2 MW (most operation at 1.5 MW)

Steady-state burnup prior ~170 MWD (a) ~750 MWD to high power pulsing

. Total number of pulses 725 158 Maximum pulse size S2.70 S3.20 Number maximum pulses 54 15 Other pulses ~80 >S2.00 ~25 >S3.00 Pulsehmeasured 520*C 550*C Pulse Tealculated(b) ~ 870* C(c) ~850' C Steady-statehmeasured 450* C(d) 550* C Steady-state Tealculated ~ 5 7 5' C( * ) ~550*C f* Additional ~100 MWD before discovery of damaged fuel.

Adiabatic at fuel OD.

d Not same element that contains thermocouples.

~400*C prior to $2.70 pulses.

(,) Higher power location compared with location of temperature measurement.

I 41

temperature caused the porous fuel structure and resultant swelling.

Pressures measured within the clad but outside the fuel matrix never exceed 40 psi.

O D

0 e

42

4. DISCUSSION OF POSSIBLE MECHANISM The information presented in Section 3 shows that while other TRIGA reactor facilities have had operations very similar to those at Texas A&M, each facility in total has a unique combination of steady-state and pulsing experience for its core. Texas A&M shows a long steady-state history cou-pled with more extensive high power pulaing than any other facility. While the operating history at General Atomic has been very similar to, and in some cases more extensive than, that at Texas A&M, there are dif ferences in

. configuration, fuel temperature, operational procedures, and the number of pulses that may be significant to fuel damage. A possible mechanism for

, fuel damage (other than direct attainment of temperatures of ~1150*C) that fits the general characteristics observed in the damaged fuel is described in the remainder of this section.

The postulated mechanism of swelling and bowing in the damaged fuel elements may be discussed as a number of interrelated phenomena:

1. Hydrogen migration under thermal gradients from regions of higher temperature to regions of lower temperature in the fuel is a governing f actor in causing swelling of the fuel.
2. Generation of high local gas pressures in the fuel matrix during very high power pulsing results in swelling and increased pore s iz e. The gas pressure is produced by hydrogen evolution in the hot spots near the surface of the fuel rod where increased hydrogen concentrations exist. These increased hydrogen concen-trations result f rom relatively long-term steady-state operation, during which the hydrogen tends to redistribute radially and axi-ally by migrating to the cooler. surface regions. However, there is a threshold temperature just above the temperatures in the thin 43

region immediately adjacent to the surface of the fuel that would ,

limit the hydrogen migration to some distance below the surface..

The higher concentration of hydrogen in the subsurface region would lead to much higher internal gas pressures at the ho* spot during pulse operation than would occur with the nominal hydride composition (ZrHl .6). To produce pressures equivalent to those in ZrHl .6 at 1150*C, the H/Zr ratio would have to increase to about 1.85 and be subjected to a temperature of about 880*C.

If the H/Zr ratio were 1.75 the required temperature would be about 1010*C.

3. The " hydrogen-depleted" regions shown in the neutron radiographs result from the loss of hydrogen to the cooler parts of the fuel.

This hydrogen evolved from the hot spots during high power pulsing .

and appears to have been absorbed by the cooler regions, espe-cially the central zirconium rod, thereby depleting the hot spots _,

of hydrogen. Under high-power steady-state operation, the hydro-gen in the depleted hot spots would be replenished (but at a much slower rate) by diffusion in the solid state and by migration in the gas phase from neighboring regions.

4. The central axial zirconium rod in the center of the fuel element appears to be a source of stress generation under the conditions encountered in the hottest parts of the fuel. These rods can swell up to ~15% in volume upon absorption of hydrogen to give an H/Zr ratio >1.7. Under extreme swelling conditions, the initial clearance between the zirconium rod and the fuel appears to be too small and the zirconium rod will swell and press against the fuel.

It appears as if stresses were generated by the swelling zirconium rod which were large enough to crack the fuel. The neutron radio-graphs and metallography of the most damaged highest-temperature portions of the fuel element indicate complete hydriding of the central zirconium rods, whereas the low-temperature, undamaged fuel shows little hydriding in the zirconium rod. In some cases 9

44

d the expanded zirconium rod actually bonds to the fuel at points of contact.

5. The structure of the hydrogen-depleted region in the distressed fuel is shown in Fig. 17. This region apparently contains signif-icant quantities of alpha phase material formed by loss of hydro-gen from the original delta phase. It is of interest to note that the fine pores in this region are largely in the form of a maze of pores at the grain boundaries. As the original delta phase trans-forms to the alpha phase, the pores that are present in the alloy are swept to the newly formed grain boundaries of the alpha phase.

Also, the change in density upon transformation from delta to the

, denser alpha phase will favor the formation of voids which will be trapped at the new grain boundaries.

From the foregoing discussion, it is possible to ascribe the bowing and swelling phenomena observed in the damaged fuel elements to the following series of events (also see Fig. 31 for additional descriptive information):

1. Hydrogen migrates toward the cooler surface regions of the fuel in an unsymmetrical configuration governed by thermal gradients and temperatures. The migration occurs over long periods of steady-state operation. See Fig. 31(a).
2. During steady-state operation, essentially no hydrogen migrates to the immediate surface region of the fuel or to the central zircon-l ium rod (which is in the hottest part of the element) because of l

i the temperature gradient. The high central temperature forces hydrogen away from the center, and very low migration rates at the immediate fuel surface region temperature prevent hydrogen build-up in this zone. See Fig. 31(a). (There is possibly some very small degree of hydriding of the central zirconium rod during high power steady-state operation by reaction with hydrogen in the gas phase in the surrounding spaces.)

45

1 BEFORE AND AFTER INITIAL HIGH POWER AFTER MANY HIGH LONG TERM STEADY- PULSING AFTER LONG- POWER PULSES FOLLOW-STA1E OPERATION TERM STEADY-STATE ING LONG-TERM STEADY-(PRIOR TO PULSING) OPERATION STATE OPERATION (A) (B) (C)

Zr Zr Zr ROD CLAD ROD CLAD ROD CLAD 500 900 -

900 -

C~

g 400 -

  1. 700 -

700 -

h300 -

200 -

500 500 h 1.6 - -- 1.6 -

1.6 -

h AFTER 0 -

0 -

0 -

PHASE ll a+5 l 6+c l ll a+6 l5+c l ll 5+c la+6 ll t t t i a a b+e g ore Fig. 31. Oualitative graphic description of possible changes in hydrogen distribution leading to fuel damage 4 0 , 8  % f (

3. Concentration of hydrogen takes place in the fuel internal to the cool surface area. This may set up the conditions for circunfer-ential cracks near the surface, possibly during later high power pulse operation. The circumferential cracks can be postulated to occur either before or after pulsing. See Fig. 31(a) and/or (b).
4. The concomitant increase in hydrogen pressure in the hot spots during high power pulsing results in swelling and an increase in pore size. See Fig. 31(b).
5. Progressive loss of hydrogen takes place in the hot spots and is absorbed in the cooling regions of the fuel during and following the pulsing. See Fig. 31(c).
6. The unhydrided axial zirconium rod in the center of the fuel element can act as a crack initiator and stress-raiser by swelling as-it hydrides 11n situ during very high power pulses and presses against the surrounding fuel element. See Fig. 31(c).

I i

47

, , . _- m_

4 - :

5. RECOMMENDED OPERATIONAL PROCEDURES To help identify potential fuel damage probleme prior to the time that the extent of damage demands removal of the fuci element from the core, the fo110 wing operational procedures are recommended.
1. Physically inspect and measure fuel care r ally after all pulsing (ot steady-state) operation in new domains for a given reactor facility to insure etntinued successful operation. Routine fuel inspection should be on an annual basis. Inspections resulting f rom operations in new domains could even be limited to the

~

immediate core regions most affected by the new operations, such as hot spot regions. For cores that require for fuel inspection the disassembly of fuel clusters involving possible damage to fittings and threads f rom excessive handling, the annual inspec-tion could include only the central, high power region of the core, elements adjacent to water holes, and a selective sampling of icwer power elements.

2. If not precluded by experimental or structural conditions, place the' thermocouple-instrumented element in the hottest position to minimize as much as 'possible the uncertainty in the maximum temperature.
3. When pulsing performance is increased to new levels, correlate new and old performance as much as possible through period, energy, power, and temperature measurements to relate to Ak. Relative energy measurements can also be made with gold foils if successive conventional detector values do not behave as expected, or as an additional correlation.

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It is useful co note that there are numerous examples of minor bending being detected in fuel elements during an inspection but disappearing at the next inspection. This is probably due to reinserting the fuel element in the core in a different angular orientation after the fuel inspection. The removal and reinsertion of an element is highly likely to result in a new rotational orientation. this change in orientation would likely be all that is necessary to reverse any possible bending caused by long-term steady-state hydrogen migration (or other possible thermal gradient stress).

However, progressive damage of the kind experienced at Teras /.&M, and manifested by the local porosity and swelling of the fuel, is terminated

,only by reducing the high power pulsing temperature to which the fuel is subjected. Since the hydrogen pressure increases nearly exponentially with fuel temperature, small temperature changes can make a very significant dif-ference in fuel damage. In fact, this relationship makes the fuel damage mechanism act, for practical purposes, as if there were a threshold tempera-ture for damage. If damage begins to appear, a reduction in measured tem-perature as small as 20*C (~40*C in peak pulsing temperature) will reduce hydrogen pressures by about 33% and is likely to stop any progressive damage.

It also seems evident that the damage threshold was just slightly exceeded in the Texas A&M core. The calculated spread in the power geners-tion among the four damaged fuel elements was about 5% and the damage ranged from very slight to significant. Also, the maximum temperature in the Texas A&M core was calculated to be only slightly greater (~20*C) than that in the FLIP core at Ceneral Atomic, where no fuel damage has occurred.

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6. FUEL TEMPERATURE SAFETY LIMIT The fuel temperature safety limit is unaffected by the events and processes discussed in the foregoing sections because it is set by the average ll/Zr ratio in the element. The safety limit is based on the result-ing hydrogen pressure exerted on the fuel element clad, tendin; to rupture it. While redistribution of hydrogen within the fuel material can result in damaging pressures localized within the fuel matrix, the pressure exerted on the clad is determined by the overpressure exterior to the fuel matrix and is set by the average H/Zr ratio in the fuel element. When the maximum pulse size is increased by relatively small increments, large numbers of
  • l l pulses would be required for damage to become evident.

It is also pointed out that, in the very few cases of fuel damage of the type discussed in this report, there was never any danger of the clad rupture that would be necessary for fission product release.

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7. OPERATIONS AT TEXAS A&M SINCE SEPTEMBER 1976 Since the discovery of the fuel damage at Texas A&M in September 1976, g pulsing operations have been suspended but steady-state operations continued without change. Since November 1979 the core has consisted entirely of FLIP elements. Routine operations produce ~100 MW-days of burnup per year. No additional fuel damage has been observed and, in fact, the least damaged of the four bowed fuel elements has continued operation and no longer exhibits any sign of damage. .

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8. REFERENCES
1. " Status Report on Damage to FL1P Fuel During Operation of the NSCR at Texas A&M University," a letter report to the Director, Division of Reactor Licensing, U.S. Atomic Energy Commission, Washington, D.C.

20545, from John D. Randall, Director, Texas A&M University Nuclear Science Center, November 1,1976.

2. Carlson, R. G. , "TRIGA Element Meta 11ography," memorandum to R. S.

Wisner, Argonne National Laboratory-W'e st, August 13, 1980. ,

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M4Nf RAL ATOMe-GENERAL ATOMIC COMPANY P. O. BOX 81608 SAN DIEGO CALIFORNIA 92138 l

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