Text

Interim Rept, Investigation of Fuel Rod Failures in CT Yankee Reactor
	ML19341A968
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	10/30/1980
From:	Klingensmith R, Pasupathi V; Battelle Memorial Institute, COLUMBUS LABORATORIES
To:	;
Shared Package
ML19341A957	List: B10137, Forwards Interim Rept, Investigation of Fuel Rod Failures in CT Yankee Reactor; ML19341A968;
References
	BCL-585-19, NUDOCS 8101290568
	Download: ML19341A968 (51)
	v • d • e

e Docket No. 50-213 ATTACHMENT CONNECTICUT YANKEE ATOMIC POW 51 COMPANY Interim Report on Fuel Rod Failures i

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BCL-585-19 INTERIM REPORT on INVESTIGATION OF FUEL ROD FAILURES IN THE CONNECTICUT YANKEE REACTOR I

j by V. Pasupathi and R. W. Klingensmith l

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Northeast Utilities Service Company i

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Electric Power Research Institute 4

i November,1980 l

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8ATTELLE l

Columbus Laboratories 505 King Avenue I-fcklymbnJRhi@' 4321

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j TABLE OF CONTENTS a

Page i

iv SUMM ARY..............................................................

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1.0 INTRODUCTION

1 L

2.0 BACKGROUND

INFORMATION ON FUEL ROD FABRICATION AND O P E R AT I O N A L H I STO R Y................................................

3 1

2.1 Fuel Rod Design Information...................

3 1

2.2 Operational History...........

5 3.0 PRELIMINA.3Y POOLSIDE EXAMINATIONS AND ASSEMBLY SELECTION F O R H OT C E L L E X AM I N AT IO N S.........................................

6 3.1 Initial Visual E xaminations...........................................

6 3.2 Second Poolside Examination.........................................

7 3.3 Assem bly Selection Criteria...........................................

7 i

.i.0 H OT C E LL E X AM I N AT I O NS..............................

9 4.1 F uel Assembly R eceipt at B C L.........................................

9 4.2 Disassembly and Fuel Rod R emoval.....................................

9 4.3 Fuel Rod Nondestructive Examinations................................. 11 4.4 D estructive E x a min a tions............................................. 25 5.0 PRELIMINARY EVALUATION OF THE RESULTS AND CONCLUSIONS......... 38 6.0 P R O POS E D FO L LOW-O N WO R K........................................ 42 I

7. 0 R E F E R E N C E S........................................................ 43 LIST OF FIGURES j

Figure

1. lodine-131 Activity in Primary Coolant of Connecticut Yankee R eactor Du ring Cycles 7 and 8.......................................

2 Figure 2. Locations of Rods Removed from Batch 8 Assembly for Examination........ 12 Figure 3. Profilometer Scan Showing Typical Ridging Observed in Rod 062E12........ '15 i

Figure 4. Small Diameter increase Observed at 63-3/4 Inches from Rod Bottom f rom R od B ottom in R od 156 E01..................................... 16 4

Figure 5. Local Diameter and Ovality increase Observed in Unfailed Rod OE2E12.......

17 f

Figure 6. Diameter l'ncrease at the Location of Tight Axial Crack Observed in F a il ed R od 013 E 03............................................... 18 i

LIST OF FIGURES (Continued)

Page Figure 7. Examples of Ridging and Diameter increase Observed in Rod 157E01 by Linear Profilometry Scan......................................... 19 Figure 8. Eddy Current Indication at the Location of Diameter increase O bse rved in R od 156 E01........................................... 20 Figure 9. Probe Coil Eddy Current Indication at the Location of Diameter i ncrease in Rod 156 E01............................................ 22 Figure 10. Gamma Scan Data Showing Depression in Activity at the Location of the Cl adding Crack.............................................. 23 Figure 11. Photograph Showing a " Bump" on Rod 156E01 at ~35 Inches f r om R od B o tto m................................................. 24 Figure 12. Photograph Showing a " Bump"in Rod 217E02 at ~35 inches f rom R od B o t to m................................................. 24 Figure 13. Appearance of the Longer of the Two Cracks Observed in Failed Rod 013 E03 at ~25 inches from Rod Bottom............................... 26 Figure 14. Appearance of the Tight Crack in Rod 013F03 at 79 Inches from R o d B o tt o m..................................................... 26 Figure 15. Example of Ovality and Ridging and Local Diameter increase Observed i n R od 5 9 5 A 1 0................................................... 27 Figure 16. Eddy Current Indication in Batch 7 Rod 595A10 at 44-3/4 Inches........... 28 Figure 17. Gamma Scan Showing Evidence of Dish Closure in Rod 595A10............. 29 Figure 18. Appearance of incipient Cladding Crack in Rod 156E0164 Inches f rom R od B o tto m................................................. 35 Figure 19. Appearance cf the Two Cracks Shown in Figure 18....................... 36 Figure 20. Appearance of the Two Cracks in Rod 157E01 at 64 incras in Etched Condition ( Etchant - Glyceregia)..................................... 37 Figure 21. Photo Montage of incipient Cladding Crack Observed in Rod 157E01 at ~3 5-3/4 i nch es f rom R od Bo ttom.................................. 39 Figure 22. Photo Montage of Incipient Cladding Crack in Rod 216E02 24 inches f ro m R od B ot to m................................................. 40 ii

LIST OF TABL CO Page Table 1.

Connecticut Yankee Fuel Suppliers and De:ign Parameters...

4 Table 2.

Sipping Results on Batch 8 Fuel Assemblies as a Function of Burnup....

6 4'

Table 3.

Locations of Failed Fuel Rods in Batch 8 Fuel Assemblies Along With F a b rica ti o n D a t a..................................................

8 Table 4.

Identification by Fuel Rods Removed for Examination from Batch 8 Assembly H07 and the Basis for Either Selection..

10 Table 5.

Fuel Rods Removed from Fuel Assembly H07 and the Extent of Nondestructive Examinations Performed....................

13 Table 6.

S u m m a ry o f N D E D a t a..................................

30 Table 7.

Fission Gas Release and Void Volume Data Obtained on Batch 8 Fuel Rods..... 33 0

SUMMARY

An increase in coolant activity was observed during reactor cycles 7 and 8 of the Connecticut Yankee reactor. Subsequent end-of-cycle and poolside examination indicated that 36 out of the 48 batch 8 fuel assemblies contained failed fuel rods. The failures appeared unique to batch 8 assemblies. A detailed coolside examinaticn campaign was undertaken along with a review of manufacturing data. Based on an evaluation of the data, two fuel assemblies, one from batch 8 containing failed rods and the other from batch 7, were shipped to the Battelle-Columbus (BCL) hot-cell facilities for examination. At BCL, the assemblies were dismantled and selected fuel rods were removed for detailed nondestructive exam; nations - profilometry, gamma scan, eddy-current scan, and visual. Based on an evaluation of the data obtained, three fuel rods from the batch 8 assem-bly were selected for destructive examinations. The three rods were punctured to collect and ana-lyze fission gas. Results confirmed that all three rods were unfailed. Two of the three rods were later sectioned to obtain three specimens for metallographic examinations. Each specimen contained the location of a local diameter increase and a coincident eddy current indication. The specimens were ground and polished in steps to the location of the anomaly. In all three specimens, cladding cracks were observed. The cracks, originating on the outer surface, were intergranular in nature, in all three cases, a fuel chip wedged in the fuel-cladding gap was observed directly across from the cladding crack. It was tentatively concluded that high local cladding stresses, caused by the presence of the fuel chip, were the primary cause of cracking in the cladding and that the fuel rod failures may have occurred by the same mechanism. Additional work is in progress to verify this conclusion.

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1.0 INTRODUCTION

Prior to cycle 7, the Connecticut Yankee reactor, which utilizes stainless steel clad fuel rods had been operating essentially free of fuel rod failures. Near the end of cycle 7, an increase in cool-ant activity was observed. This increase continued during cycle 8 operation, as shown in Figure 1.

i Examination of the coolant ac'. vity data showed that a large portion of the activity was due to the recoil process (evidenced by low l-131/l-133 ratios) and the presence of Np239 in the coolant.

1 This examination indicated that the fuel in the f ailed rods was directly exposed to the coolant, in 4

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addition, a high Cs134/Cs 137 activity ratio indicated that the source of the activity was high burn-i up fuct (probably batch 8 fuel),(II At the end of cycle 8, all of the discharged batch 8 and selected batch 7 and batch 9 fuel as-j semblies were leak tested using a wet sipping system. The results showed that 36 out of the 48 batch 8 fuel assemblies (75 percent) were leaking, while all six batch 7 Jnd batch 9 fuel assemblies that were sipped were sound.Il) Four of the batch 7 fuel assemblies sipped had undergone three cycles of operation. Therefore, fuel rod failures appeared unique to batch 8 fuel assemblies. More recent examination, after cycle 9 operation, however, has indicated fuel rod failures in batch 9 I

assemblies.(21 After leak testing, a number of batch 8 fuel assemblies were visually examincd using an under-water television system in the spent fuel pool. In four of the assemblies, failed rods were observed I

- two rods each in three assemblies and eight rods in the fourth. All failures were in the form of axial cladding cracks of various length and widths. Some of the cracks appeared to spiral along the rod length. All failures appeared to be of a brittle nature. Multiple cracks were observed at the i

same axial location on some of the rods.

While the data obtained from this initial poolside examination indicated the nature of fuel rod failures, the cause of failure could not be identified. A review of batch 8 fabrication data and an i

analysis of reactor operational data for cycles 7 and 8 were perforrred. Based on the results of these efforts, it was postulated that the failures could have been cause.d by a power ramp near the end of cycle 7 or by a " wear-out" or " life-limiting" mechanism. In any case, further investigations, j

I including hot cell examination, were deemed necessary to clearly identify the failure mechanism.

As an interim measure, restrictions on the rate of power increase for the Connecticut Yankee reactor were imposed to prevent failures in future operating cycles.

A more detailed poolside examination was undertaken leading to the selection of two fuel

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j assemblies, one batch 7 and one batch 8, for hot cell examinations. Both were three-cycle assem-blies. The batch 8 assembly contained several visually observed failed fuel rods. The objectives of the hot cell examinations were to:

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Identify the cause of fuel rod failures in batch 8 2 uel differences, if any, between batches 7 and 8 could have 2.

Establish what UO f

contributed to the failures 3.

Identify changes required in fuel rod design, reactor operation, or both to avoid failures in future cycles.

This interim report is intended to document background information, results from preliminary poolside examinations, and the results from the initial phase of the hot cell examinations. High-lights of results, preliminary conclusions, and proposed follow on work are identified. Detailed evaluation of the results has been deferred pending completion of the next phase of hot cell exam-inations and will be documented in the final report.

2.0 BACKGROUND

INFORMATION ON FUEL ROD FABRICATION AND OPERATIONAL HISTORY 2.1 Fuel Rod Design Information Fuel rod design for the Connecticut Yankee reactor had changed very little over the years with 2 uel pellet fabrication techniques and suppliers. For the first six batches, fuel the exception of UO f rods were designed and fabricated by Westinghouse. For batch 7, the fuel pellets and the rods were fabricated by Gulf-United Nuclear Fuels (GUNF) Batch 8 fuel pellets were fabricated by British Nuclear Fuels Limited (BNFL) and assembled into rods by Gulf-United Nuclear and Babcock and Wilcox Company (B&W). Fuel rods for batches 9 through 12 were fabricated by Babcock and Wilcox. The cold-worked type 304 stainfess steel cladding (weld drawn) for all batches was sup-plied by Superior Tube Co. Table 1 summarizes various fuel vendors and some of the design i

parameters.

i Since fuel rod failures in large numbers (60 rods or more) initially appeared unique to batch 8 -

fuel assemblies and also in light of very good fuel performance in prior batches, it is important to determine what, if any, differences exist in batch 8 fuel relative to earlier batches. Northeast Utilities and Babcock and 'Wilcox have performed a detailed review of available manufacturing information, specifications, QA records, and as-built data. In addition, examination of archive pellet and cladding samples from batches 7,8, and 9 were performed by B&W.(3)

Examination of as-built data and manufacturing information on the pellets showed batch 8 l

pellets to be somewhat different from those of batches 7 and 9. The batch 8 pellets were manu-factured by BNFL using a controlled porosity (CONPOR) process. However, no problems or deficiencies relating to fuel rod failures could be identified from the manufacturing data. It was also found that all UO pellet and fuel rod nominal dimensions and parameters, as well as specifi-2 cation requirements for all three batches were essentially the same.III

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TABLE 1. CONNECTICUT YANKEE FUEL SUPPLIEns AND DESIGN PARAMETERS 6fy i

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9 Batch Designation

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3 4

4N 4A 5/7A 5G SA

' Fuel contractor

' Westenghouse (W)

NUMEC NUMEC W

GGA GGA W

GUNF Fuel pellet supplier Westinghouse (W).

NUMEC NUMEC W

BNFL BNFL

.W GUNF BNFL B&W Fuel clad supplear Superior Tube Company

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Suswr nor Fuel assembly fabricator Westinghouse (W)

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50h 51/2 48 56 Number of assemidies.

53 52 52' 48 2

2 49/1 1

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. Enrichment, w/o 3.00 3.24 3.67 3.00 4.00 4.00 3.25 4 0/3.67 4.0/3.66 4.0 D-Fuel density, % '

93.0 94.0 94.0 92.8 93.2 92.4.

92.9 93.2 93.0 92.7/92.8 94.85/94.57 95.17 95 27 3

initial pressure, psia 14.7 Fdi ges Au

He He Air He He Air He/A / Air r

1 Stack height. in.

121.8 121.8 121.8 120.0 119.3 118.4 120.0 121.5 121.1 120 0 120.3 D-Pellet diameter,in.

.0.3835 03680 03835

-03645 03835 Po'let length, in.

0.600

- 0.450

- 0.600 0.420 c

Cladding snat'l SS SS SS' SS SS lirc SS SS Zuc SS SS j

Clad thickness,in. ~

0.0165 0 024 0.165

- 0I)25 0.165 Clad t.D.. in.

0.389

0.374 0389

> 0.3735 0.389 Gap, deemster, in.

0.0055 0.006 0.0055

> 0.009 0 0055 e

Four assemblies contairi four incalloy test rods each.

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5 The investigations conducted by B&W on archive pellets also showed some differences be.

m batch 8 pellets and those from batches 7 and 9. The examinations performed included metallog-raphy, pore size distribution measurements, and thermal resintering tests. Batch 8 fuel pellets, in comparison to batch 7 and 9 pellets, appeared to contain a higher percentage of larger pores in a high-density rnatrix, orobably because of the use of a pore former during fabrication. Quantitative analysis of the photographs (on the Quantimet) also confirmed this observation. In batch 8 pellets, 50-60 percent of the porosity was in the 5-10 gm range, while in pellets from batches 7 and 9, 70-80 percent of the porosity was in the 1-5 gm range.

Thermal resintering tests were performed on pellet samples from batches 8 and 9. Tests were conducted for 6 and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 1600 C and 1700 C. Results showed no significant difference in the measured density changes. For batch 8 fuel, the increase in density, after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 1700 C, ranged from 1.18 to L43 percent. For batch 9, it was 1.43 percent.

In summary, batch 8 fuel pellets do appear somewhat different from batch 7 and 9 fuel pel-lets by virtue of the fabrication process. The larger fraction of small pores (1-5 pm) observed in pellets from batches 7 and 9, compared to batch 8 pellets, suggests that fuel from the former may be more densifying in the reactor than is the batch 8 fuel. One aspect of lower in-reactor densifi-cation is that it results in fuel-cladding contact earlier in the assembly's life and therefore relatively higher cladding stresses at higher burnups. B&W also examined cladding archive samples. No dif-ference of any significance was observed.

2.2 Operational History A review 1e power history of batch 8 and 9 fuel rods was performed by NUSCO. Average linear H. sting of batch 8 fuel throug5 cycle 7 was approximately 6.6 kw/f t. During cycle 8, it was 5.5 kw/f t. Batch 9 fuel rods experienced approxirnately 6.5 kw/ft during the second cycle of irradiation (cycle 8). Thus, the power history of batch 8 and 9 fuel rods appears quite similar.

An increase in coolant ectivity was observed during the second cycle of batch 8 fuel rods. The average burnup of these rods at the end of the cycle was approximately 22,500 MWD /MTU. The l

average burnup of batch 9 fuel rods at the end of their second cycle was 23,900 MWD /MTU (ap-proximately 6 percent higher), yet no failures were evident (based on the limited examinations) in batch 9 fuel assemblies. Tht's, it appears that no clear correlation exists between power histories and the presence of leakers. This may also indicate that failures in batch 8 occurred under normal operating conditions.

However, examination of sipping results indicates a definite correlation between burnup and leakers, as shown in Table 2.(1) At an assembly burnup of 34,200 MWD /MTU and above, all as-semblies (a total of 20) were found to be leaking. Between burnups of 32,400 and 32,500 MWD /

l MTU,13 out of the 16 assemblies were leaking. At burnups below 32,400 MWD /MTU, only 3 out of 12 assemblies were found to be leaking.

6 TABLE 2. SIPPING RESULTS ON BATCH 8 FUEL ASSEMBLIES AS A FUNCTION OF BURNUP Assembly Burnup, MWD /MTU Lesking Sound 35,800 H07,H21,H31,H38 35,000 H03,H23,H40,H53 34,300 H 16,H20,H33,H35 34,200 H05,H 13,H 15,H24,H30,H37,H43,H44 32,500 H 10,H 14,H22,H 28,H29,H32 H45,H50 32,400 H 11,H 18,H25,H27,H41,H45,H 52 H36 31,500 H26, H34 H04,H06,H 17,H39,H42,H 51 30,100 H12,H19,H49 29,800 H01 36 Leaking 12 Sound The operating history of batch 8 fuel was also examined for any departures from steady-state full-power operation. The most significant departure was a 10-day period at 65 percent of full power near the end of cycle 7 (near the end of second cycle for batch 8 fuel). The reactor power was subsequently raised to 100 percent. It has been hypothesized that this power ramp could have caused or initiated the failures in batch 8 fuel by inducing excessive local clad strains adjacent to radial pellet cracks. It was, however, concluded that additional efforts were needed to either substantiate or refute this hypothesis.

3.0 PRELIMINARY POOLSIDE EXAMINATIONS AND ASSEMBLY SELECTION FOR HOT CELL EXAMINATIONS 3.1 Initial Visual Examinations At the conclusion of the sipping campaign,24 fuel assemblies were visually examined using

, underwater television equipment in the spent fuel pool.' in four fuel assemblies, failed fuel rods were observed - three assemblies with two failed rods each and the fourth with eight failed rods.

7 All four assemblies had been classified as leakers from the sipping examination. All failures were in i

the form of axial cladding cracks of varying lenghts and widths. Some of the cracks appeared to spiral along the rod length. Multiple cracks were observed in some rods at the same axial location.

All failures appeared to be of a brittle nature, t

3.2 Second Poolside Examination While the initial examination campaign at the reactor site was to explore the nature of failure, a second campaign was undertaken to characterize cladding damage and identification of failed rods, and to select fuel assemblies for hot cell examination. Prior to this campaign, a detailed review of I

preirradiation data on candidate assemblies was conducted. Data of major interest were those that relate to the propensity for early pellet-clad gap closure. Data reviewed included (1) mean density and standard deviation of each pellet lot, (2) fuel densification under thermal resintering tests, and (3) clad tube lot inner diameters.

i Along with the above review, six batch 8 fuel assemblies (all leakers) were examined by Combustion Engineering in the spent fuel pool using a periscope. From this,29 failed fuel rods were identified as failed. Table 3 shows the results of the examination along with the fuel and clad lot identifications, burnups of the failed rods, and initial weight of fuel. The weight of fuel in the rods is significant since it indicates average initial fuel density and thus provides a measure of the extent of possible in-reactor densification.

1 3.3 Assembly Selection Criteria Based on the review of the data, batch 8 fuel assembly HO7 was selected as the best candidate for hot cell examination of an assembly containing failed fuel rods. The reasonina behind the selection of this assembly is:

HO7 is in the highest burnup assembly group and has no grossly damaged peripheral i

i rods that could complicate later handling.

HO7 contains a broad representation of the batch 8 pellet and cladding lots, l

including combinations with a relatively high propensity for early gap closure.

HO7 contains the only rod observed with a single cladding crack that may be the e

primary failure and, therefore, a backup for the hot cell scarch for incipient evidence of failure,i.e., sound rods with nonperforating cladding cracks.

e HO7 has seven failed rods identified in the visual examination. (The probability of finding evidence of incipient failure in this assembly is believed to be at least as high as for any other assembly in the highest burnup group.)

HO7 contains cladding lots also used in the batch 7 assembly selected for.

examination.

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8 TABLE 3. LOCATIONS OF FAILED FUEL RODS IN BATCH 8 FUEL ASSEMBLIES ALONG WITH FABRICATION DATA Estimated Assembly Rod Rod Tube Pellet UO2 Burnup at EOC-8, Number Location (a)

Number Lot Lot Wt, g Gwd/MTU H07 K15 179E09 103 4

2,299 35.1 J13 118E08 129 9

2,288 36.5 J14 356E09 109 14 2,297 35.8 H13 130E06 24 12 2,282 36.5 P13 346E03 109 10 2,299 35.6 E14 077E04 103 8

2,287 35.3 A1 013E03 28 11 2,283 35.0 H31 M13 900E01 117 5

2,293 36.3 E14 823E07 118 5

2,293 35.7 K1 823E03 118 5

2,296 35.1 H38 A3 081E01 129 12 2,288 34.9 A5 984E07 124 12 2,293 34.9-83 526E06 114 12 2,287 35.6 H33 D14 496E08 112 7

2,296 34.3 P14 488E02 112 7

2,298 33.3 N14 496E10 112 7

2,296 34.1 N9@

595E06 113 3

2,301 34.9 K10(b) 475E05 111 7

2,296 34.3 l

K9(b) 483E01 112 7

2,289 34.3 H16 M3 233E12 102 4

2,297 34.9 I

A8 251E03 102 4

2,300 33.5 A10 524E05 114 3

2,295 33.5 d

C14 208E03 101 3

2,299 34.0

.i14 254E02 102 4

2,298 34.0 H53 MIS 448E08 111 2

2,297 34.5 B15 455E05 111 2

2,301 33.6 P12 450E08 111 2

2,296 35.3 r

N9 238E02 102 1

2,296 36.0 G3 838E01 116 1

'2,296 35.1 E2 799E08 121 1

2,297' 34.4 l

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(a) Refer to Figure 2 for fuet rod location scheme.

(b) oue to depth in assembly, unsure Of l0 Cation Of f ailed rod, DO5Sibly 2 out of 3 faded.

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in addition to the batch 8 fuel assembly, one batch 7 fuel assembly ( Assembly No. G11) was examined. This assembly was from the highest burnup group (36,700 MWD /MTU). It was previ-ously confirme ' as sound by sipping. Visual examination of the assembly showed no significant anomalies. Other reasons for selecting this assembly where: (1) the presence of pellet lots previ-ously characterized with respect to simulated fuel densification behavior so that in-reactor densi-fication and a comparison of batch 8 fuel behavior can be obtained and (2) one such characterized pellet lot is in the highest burnup rod group, as well as in several rods with clad tube lot 23, which is also used in rods of the batch 8 assembly HO7.

4.0 HOT CELL EXAMINATIONS 4.1 Fuel Assembly Receipt at BCL Fuel assembhes HO7 and G11 were shipped dry to BCL. In each case, the cask was unloaded using routine cask handling procedures. However, during unloading of HO7, release of radioactive contamination into the unloading work area resulted in significant contamination of the facility.

During transit, individual rods in the fuel assembly apparently became heated to temperatures of approximately 500-550 F. It is believed that this temperature history had no adverse affect on the unfaiN rods in the assembly, however.

No problems were experienced in the unloading of fue; assembly Gil, which contained no failed fuel rods and had cooled for 30 months compared to 15 months for HO7.

4.2 Disassembly and Fuel Rod Removal The fuel assemblies were transferred into the hot cell and disassembled at the top end by re-moving the upper nozzle to facilitate removal of fuel rods. No problems were encountered during removal of the upper nozzle, initially, two rods were removed from assembly HO7. These were peripheral rods previously identified as failed during poolside visual examination. These rods were removed to verify assembly orientation and location of unfailed candidate rods for examination. Also, one of these failed rods appeared to have only one crack, which was believed to be O primary defect, and therefore of in-terest in determining the cause of failure. Subsequently, X ihional unfailed rods were removed and subjected to detailed examinations. Selection of these *.t Js was based primarily on fuel rod burnup. Other bases for selection included pellet and tube lot characteristics and proximity to failed fuel rods. Table 4 shows the rods removed and the basis for their selection.

TABLE E-4. IDENTIFICATION OF FUEL RODS REMOVED FOR EXAMINATION FROM BATCH 8 ASSEMBLY H07 AND THE

- BASIS FOR THEIR SELECTION Estimated UO2 Rod Pellet Stack Tube

Burnup, Number Location (a)

Lot Wt, g Lot Gwd/T Basis for Selection 013E03 A1 11 2,283 28 35.0 Rod with single crack (failed rod) 062E12 L8 11 2,280 23 36.7 2nd highest BU with same tube lot as in assembly G-11 065E11 G4 11 2,286 25 36.8 Highest BU group in fuel assembly g

G13E06 F4 11 2,286 28 36.8 Highest BU group in fuel assembly 157E01

- M12 10 2,300 129 36.7 2nd highest BU group in assembly ]

j Proximity 010E03 L12 11 2,286 30 36.7 2nd highest BU group in assembly > to failed i

rods 075E05-K12 8

2,286 103 36.7 2nd highest BU group in assembly J 217E02 J12 9

2,288 101 36.7 2nd highest BU group in assembly; pellets in lots 1-10, small clad ID 1007E01

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'14 2,295 123 36.8 Highest BU group in F/A,large pellet density std. dev., small clad ID (t) See Figure 2 for rod array in assembly H07.

11 Af ter the selected unfailed rods were removed. attempts were made to remove the other five known failed fuel rods in the assembly. The objective here was to eliminate release of activity from the failed rods into the BCL transfer pool. Removal of these rods was extremely difficult because of significant diameter increases (presumably caused by cladding splits). While four of the rods were removed with difficulty, the fifth rod could not be removed. The roa eventually broke ap-proximately 10 inches from the top end. The remainder of the rod was pushed back into the assembly. Figure 2 shows the location of rods removed from this assembly.

Four fuel rods were removed from assembly G11. These rods had a range of burnups, with two rods in the highest burnup group, in addition, all four rods contained fuel pellets from lot 512, for which thermal resintering data were available.

4.3 Fuel Rod Nondestructive Examinations 4.3.1 Fuel Rods from Batch 8 Assembly HO7. All the unfailed rods and one of the failed fuel rods removed were subjected to detailed nondestructive examinations (NDE). These examinations in-cluded profilometry, gamma scan, eddy current scans (encircling coil and probe coil), and stereo-visual examinations. Table 5 shows the extent of NDE performed on the rods removed from the fuel assembly.

4.3.1.1 Profilometry. Fuel rod profi'ometry was performed to detect and characterize any tocal fuel rod diameter anomaly and to determine the extent of ovality and cladding creepdown over the full length of the fuel rod. The Battelle profilometer used for this work consists of two linear variable' differential transducers (LVDT) placed 120 apart, which move axially along the fuel rod. The LVDT's are mounted on a carriage assembly, which is attached to a support structure.

The carriage assembly moves up and down the support structure. The LVDT's are mounted so that the transducer shafts float in an air chamber. The carriage assenibly also contains pairs of guide rollers located above and below the transducers to hold the rod in position during profiling. As the LVDT sensing heads traverse axially along the fuel rods, the transducers respond to changes in the fuel rod diameter. These responses are amplified and converted electronically into strip chart recorder signals. T ne profilometer is also equipped with an event marker, which marks the strip chart for every 0.1 inch of the rod profiled. The axial location of any measurement is known within 10.05 inch. The fuel rod may be profiled in a rotating or nonrotating mode.

The profilometer is calibrated with a step standard, which is a cylindrical rod with precisely machined steps of different diameters.

Nine rods were profiled. All rods (failed and unfailed) showed significant cladding ridging at fuel pellet interfaces. Typical ridge height (relative fuel rod diameter at mid pellet region) was

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R South Failed rods

__J Rod removed for examinations FIGURE 2. LOCATIONS OF RODS REMOVED FROM BATCH 8 ASSEMBLY

13 TABLE 5. FUEL RODS REMOVED FROM FUEL ASSEMBLY H07 AND THE EXTENT OF NONDESTRUCTIVE EXAMINATIONS PERFORMED NDE Known Coil Probe Detailed F.G.

Rod No.

Position Failed Profile E.C.

E.C.

Visual Punch 7-Scan Comment Assy H-07 013E03 Al' Yes X

X X'

X Removed 179E09 K15 Yes Removed 062E12 L8 No X

X X

X Candidate rod 065E11 G4 No X

Reinserted 013E06 F4 No X

X Removed 157E01 M12 No X

X X

Candidate rod 010E03.

L12 No X

X X

R e oved 075E05 K12 No X

'X Reinserted 217E02 J12 No X

X X

X Candidate rod 1007E1 E4 No '

X X

Removed 130E06 H13 Yes Removed 077E04 E14 Yes Removed 346E03 P13 Yes Removed 356E09 J14-Yes Remtved 118E08 J13 Yes Broken 3

I 14 approximately 1 mil. Fuel r3d ovality was rather low,less than about 4 mils at the top and bottom ends and about 0.5 to 1 mil in the mid regions. Fuel rod average creepdown based on the diameter measured in the plenum was in the range of 1 to 1.5 mils.

In addition to clad ridging, some of the unfailed rods showed local diameter increases of up to 5 mils. The failed rods exhibited large diameter increa:es at the locations of cre:ks obrsnied on the rods.

Figures 3,4,5, and 6 show examples of ridging and diameter increases in two unfailed rods and one failed rod, respectively.

Selected regions of three of the unfailed rods showing local diameter increases were also pro-filed in the nonrotating mode to produce linear scans. These scans were made at 30' intervals. The objective was to determine whether increases were circumferential or localized at one small spot, in all cases, the diameter increases were caused by localized spots sr,anning no more than 30 of the rod circumference. The data also showed that the ridges observed extended uniformly over the fuel rod circumference. Figure 7 shows appearance of the ridges and diameter increases in rod 157E01, 4.3.1.2 Eddy Current Scans. The selected fuel rods were scanned using both the encircling coil and probe coil eddy current systems. Selection of fuel rods or regions of rods for the probe coil scans was based on an evaluation of the results from encircling coil scans.

Encircling coil eddy current inspection involves excitation of a small annular shaped coil of insulated copper wire with an alternating electrical current to induce an alternating magnetic flux in the vicinity of the coil windings. The alternating flux penetrates the cladding when the fuel rod is inserted in the coil and induces currents that tend to flow in a circumferential direction in the clad-ding around the central axis of the fuel rod. The magnitude and phase of these currents is reflected in the complex electrical impedance seen at the terminals of the test coil. The eddy current instru-ment translates test coil impedances into output voltages. Defects in the cladding (e.g., cracks) disrupt the flow of the eddy currents, changing the measured impedance of the test coil. Defects are detected and identified by changes in the horizontal output (e.g., in-phase component) and the vertical output voltages (e.g., quadrature component) that are recorded on the chart.

As a qualification scan a " standard tube" containing machined defects was tested.

Results of the encircling coil scans showed that the eddy current system was highly responsive to the clad ridging. In addition to the ridge signals, a number of defect indications were observed.

The locations of these indications coincided with those of fuel rod diameter increases observed by profilometry. Figure 8 shows a typical example of an eddy current indication at the location of diameter increases.

Selected regions of several fuel rods were also scanned using the probe coil eddy current system. These regions had shown one or more defect indications in the encircling coil scans.

15

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48 49 50 Axial Location, inches from rod bottom FIGURE 3. PROFILOMETER SCAN SHOWING TYPICAL RIDG:NG OBSERVED IN ROD 062-E12

16 Local Diameter increase 0.424 g

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60 61 62 63 64 65 Axial Location, inches from rod bottom FIGURE 4. SMALL DIAMETER INCREASE OBSERVED AT 63-3/4 IN.

FROM ROD BOTTOM IN ROD 157E01

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39 40 41 42 Axial Location, inches from rod bottom FIGURE 5. LOCAL DIAMETER AND OVALITY INCREASE OBSERVED IN UNFAILED ROD 062E12

18 0.428 --_- -- --

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T Axial Location, inches from rod bottom FIGURE 6. DIAMETER INCREASE AT THE LOCATION OF TIGHT AXlAL CRACK OBSERVED IN FAILED ROD 013E03

19 23.7 4

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23 24 Axial Location, inches from rod bottom FIGURE 7. EXAMPLES OF RIDGING AND DIAMETER INCREASE OBSERVED -

IN ROD 157E01 BY LINEAR PROFILOMETRY SCAN

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L FIGURE 8. EDDY CURRENT INDICATION AT THE LOCATION OF D1AMETER INCREASE OBSERVED IN ROD 157E01

o 21 The eddy current test using the localized probe coil offers improved sensitivity in the location of incipient defects when compared to the encircling coil technige The test involves the simulta-neous excitation of a contoured probe coil at 100 KHz and at 700 KHz. The resultant alternating flux penetrates the cladding in the vicinity of the probe coil and induces current to flow within a confined volume of the cladding. Tha magnitude and phase of these currents are reflected in the complex electrical voltages seen at the terminals of the test coils. The two-frequency eddy current instrument separates the valtage response from each of the two frequencies,100 KHz and 700 KHz, and provides two detected output voltages.

The outputs are recorded on two channels of the chart recorder. A third channel records the summation of the output signals produced by 100 KHz and 700 KHz detectors. This signal combi-nation enhances the defect signals while minimizing signals caused by unwanted variables (e.g.,

probe wobble).

When a fuel rod is being inspected, the probe is moved axially along the surf ace of the fuel rod. The probe inspects a circumferential segment which is slightly greater than 50 degrees. After the first pass, the rod is rotated 45 degrees to again scan axially along the rod surface. Thus, com-plete coverage of the rod circumference is accomplished in 8 passes.

Results from probe coil scans showed much stronger indications at the same location as en-circling coil scans. In addition, the indications were also found to be localized, i.e., restricted to a small fraction of the fuel rod circumference. This observation confirms those from the profilometry scans. Figure 9 shows a typical example of a probe coil indication.

4.3.1.3 Gamma Scans. Selected unfailed and failed fuel rods, as indicated in Table 5, were gamma scanned r ver the length of the fuel stack. The roos were scanned for both gross gamma and Cs137 activity profiles. The results in general snowed no anomatics. Activity depressions at pellet interf aces were clearly observed indicating little or no pailet dish filling or axial redistribution of fission products. In the case of the failed fuel rod, small reductions in activity were observed at the locations of cracks, as shown in Figure 10. This was presumably due to the teaching of fuel out of the rod.

4.3.1.4 Visual Examinations. Selected rods were examined using a stereovisual examination system. Emphasis was placed on the regions of rods where anomalies had been identified by other NDE data. In two of the unfailed fuel rods, visible " bumps" were observed at the locations of di-ameter increases and eddy current indications. Figures 11 and 12 show the appearance of these bumps in two of the rods.

The unfailed rods, in general, appeared to be in excclient condition with very little crud present. Crud decoration of the ridging on the cladding could, however, be readily seen.

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25 oo Only one of the failed rods was visually s.ined in detail This rod had been previously identified as having a single crack probably resulting f rom a primary defect. Upon closer exami-nation in the hot cell, a second crack was observed approximately 50 inches away from the other crack. The two cracks appeared to be unrelated to eacn other. Locations of these crmAs corres-ponded to the locations of anomalies observed in the other NDE data. Figures 13 anc 14 show the appearance of the two cracks in this rod.

4.3.2 Fuel Rods from Batch 7 Assembly G11. Two of tN f our fuel rods removed from assembly Gil (ba,tch 7) were subjected to profilometry, eddy current scans, and gamma scans. Profile traces for the batch 7 rods appeared significantly different than thov of the batch 8 rods. This can be seen by comparing Figures 3 ar.d 15. Even so, the profile characteristics were basically similar.

Both batches showed a similar degree of ridging, ovahty, and creepdown.

In each case, the ridge heights were similar. Batch 7 rods showed slightly higher ovality (about 3 mils in the rod mid-region) and slightly more creepdown (about 2 mils) than did HO7 rods. In one of the rods a small local diameter increase was observed. A typical example of rod ovality and ridging and local diameter increase is shown in Figure 15.

Eddy current scan (encircling coil) results were very similar to those from HO7 rods but with very few indications. Figure 16 shows one of the eddy current indications observed.

Results from the gamma scan of the two rods showed somewhat different characteristics in comparison to the HO7 fuel rods. The mid-region of the rod showed no activity depres:; ions at pellet interfaces (see Figure 17). A possible explanation is that the fuel in these rods was more plastic (at power) or had swollen more than that in HO7 rods resulting in closure of the dishes.

4.3.3 Summary of NDE Results. Table 6 summarizes the results from all of the nondestructive exammations performed on the rods. The data show highly consistent results from the different NDE proc ~dures, with good correlation of locations of the anoma!ies. Examination of all the data together also provides a sound basis for selecting rods and rod sections for destructive examination.

Based on a review of data, three unfailed batch 8 fuel rods were selected as prime candidates for incipient crack searcn by destructive examination. These rods were 157E01,217E02, and 062E12. All three rods contained eddy current indications at locations of diameter increases and visible bumps.

4.4 Destructive Examinations The primary objective of the destructive examinations was to confirm s.e existence of incip-ient cracks related to the primary clad breaching mechanism, io characterize those cracks, and to

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FIGURE 17. GAMMA SCAN SHOWING EVIDENCE OF DISH CLOSURE IN ROD 595-A10.

(Note lack of activity depression at pellet interfaces compared to Figure 10.)

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Eddy Current (b)

Rod No.

Profilometry Data Encirchng Probe Gamma Scan Visual E nam Status 010 E03 Diam increase at 17 ", 26 ", 3 2",

Weak indications Ridge type NP NP Returned to asseinoly L12 48 1/2",78 1/2",82" and. cation at Ridging 17",32" E Cl at 26" 481/2" 51 1/2".53" 56",57",102-1/4 ", 84", 86" 1007 El Ridging Few weak NP NP NP in storage E4 Ovality change at 78",110 1/2" indications 065 E11 Ihiging NP NP NP NP Returned to assembly 3

G4 Ovahty change at 11-1/2",61-1/2", 86" 075-E05 Ridging Many weak NP NP NP Returned to assembly K12 Small diam increase 371/2",47",

indications 67",85-1/2",97" Fuet / ssembly G-11 595-A12 Larger ovality than H07 rods Wi:.k ECl at ~ 8 NP Noisier gross gamma trace than NP in stora9e F11 Rn!9 ng diam increase at 24-ECl at ~ 42" It07 rods Few activity spikes 1/2", 38" 535-A10 R niging Strong ECl at approm NP Same as 595-A12 NP in storage Larger ovality than H07 rod 44" Diam increase at 44-1/4"

32 determine the mechanism of failure. This was to be accomplished by metallographic examinations.

The other important objective of the destructive examinations was to provide background informa-tion on general fuel rod performance. Destructive examinations conducted thus far have been limited to batch 8 fuel rods from assembly HO7.

The destructive examinations on the selected rods consisted of:

(1) Fuel rod puncture for fission gas collection and fuel rod internal volume measurement (2) Fuel-clad metallography.

4.4.1 Fission Gas Collection and Analysis, and Void Volume Measurement. The objectives of performing fission gas collection and analysis were to determine if the fuel rods were unfailed and to determine the fractional fission gas release during irradiation.

The three selected fuel rods were punctured, released gases were collected, and the volumes of release were measured using the BCL Hot Laboratory fission gas collection system. The fission gas collection system consists of a fuel rod punch chamber, a series of calibrated expansion volumes containing two pressure gauges (0-15 psi and 0-100 psi), a gas collection system with a Toepler pump, and an evacuation system. The procedure used for fuel rod puncture, released gas volume measurement, and gas collection was:

(1) The fuel rod was inserted through the gas-tight punch chamber and sealed with teflon seats and nuts at each end of the chamber.

(2) The entire system - punch chamber, expansion system, pressure gauges, and collection system - was evacuated.

(3) The prepunch system volume was calibrated by pressurizing the punch chamber to about 80 psi with helium and expanding into a known volume. The helium pressures before and after expansion were noted. The system volume was then calculated by the amount of pressure drop. This step was repeated 10 times.

(4) Readings from the prassure gauge were recorded at the time of puncture, at 2 minutes, and, subsequently, at 5-minute intervals until the system equilibrated.

(5) The rod gas was then collected in glass bulbs for spectrometric analysis.

(6) The fuel rod void volume was determined by backfilling the punch chamber and -

the fuel rod to about 80 psi with helium and expanding into the calibrated volume. This procedure was repeated 10 times.

Table 7 shows the gas release and void volume data obtained on the three fuel rods. Also shown is the estimated volume of fission gas generated and the fractional fission gas release obtained.

Data for the three rods agree closely. The data confirmed that all three rods were unfailed. The fractional releases observed are believed to be in the range expected in a typical commercial PWR

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34 fuel rod. The low fractional releases observed indicate that the temperatures experienced by the fuel in these rods were not high enough to cause significant UO2 pellet grain growth or restructuring, which would result in redistribution and release of fission products. This observation is consistent with the conclusions reached from the Cs137 gamma scan data.

4.4.2 Metallographic Examinations. From an evaluation of the NDE data (profilometry, eddy current, and visual examinations) tnree locations from two of the rods were selected for the initial phase of metallographic examinations. These locations were 64 inches and 35-3/4 inches from the bottom of rod 157E01 and 24 inches from the bottom of rod 217E02. All three locations had shown eddy current indications of local diameter increases (see Table 6 and Figures 4,8, and 9).

4.4.2.1 Specimen Preparation and Examination. The specimens were cut from the rods using a water-cooled abrasive wheel. The cuts were made approximately 1/4 inch away from the target of indications. The samples were enclosed in a stainless steel supporting tube and then held in Bakelite premounts by means of a cold setting ecoxy resin. The azimuthal orientation of the sample was maintained by aligning an indentation on the sample with a circular indentation ma-chined on the Bakelite mount cavity.

Generally, the specimens were ground and polished incrementally to permit examination of many surf aces. Progress through the specimens was followed by measuring the mount length with a micrometer when each new surface was reached. Grinding at each level was performed with suc-cessively finer silicon carbide abrasive papers through 600 grit using water as a lubricant. This was followed by polishing with a 1-g alumina suspension in a 2 percent chromic acid solution. Each surface was examined both in the as-polished condition and after being etched with glyceregia.

4.4.2.2 Metallographic Results. Cracks were found in the cladding in all three specimens. The cracks were intergranular in nature and started at the outer cladding surface. Details of the results on each specimen are given below.

In the specimen taken 64 inches from the bottom of rod 157E01, a crack initiating between the sixth cod seventh surfaces was found, but it disappeared between the ninth and tenth surfaces.

The crack was associated with a fuel chip wedged in the fuel-cladding gap and a " bump" in the cladding. Additional cracks were observed on the eighth surface. All cracks were intergranular, originating on the outer surface. No cracks were observed on the inner surface. Total crack length was approximately 0.125 inch, with the maximum depth of penetration bein approximately 70 per-cent. Figures 18,19, and 20 snow the appea nce of the crack.

No fuel restructuring was apparent. There was no visible fuel-cladding gap, except in the vicinity of the fuel chip. The fuel crack pattern shows circumferential cracking near the pellet periphery. No radial fuel pellet cracks adjacent to cladding cracks were found.

35 Backup Ring

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i 35X HC49988 and 49989 FIGURE 18. APPEARANCE OF INCIPIENT CLADDING CRACK IN ROD 157E01 64 INCHES FROM ROD BOTTOM i

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FIGURE 20. APPEARANCE OF THE TWO t, RACKS IN ROD 157E01 AT 64 INCHES IN ETCHED CONDITION (ETCHANT - GLYCEREGIA) if

s 38 Seven surfaces were examined on the specimen taken at 35-3/4 inches from the bottom of rod 157E01. Cracks were observed on the sixth and seventh surfaces. One major crack (approximately 50 percent penetration) and many small cracks were observed. Further grinding of the specimen was stopped. All cracks were intergranular, originating on the outer surfaces. Again, the cracks were associated with a fuel chip in the gap and a bump" in the cladding. Figure 21 is a photo montage of the crack.

Eight surfaces were examined on the specimen taken 24 inches from the bottom of rod 217E02.

A number of small cracks were first seen at the sixth surface. Two major cracks (largest approximately 75 percent of wall) and a few small ones were observed. These cracks also were associated with a fuel chip in the gap. All cracks were intergranular, originating at the outer cladding surface. Figure 22 is a montage of the crack.

Results of the metallographic examination of the three specimens can be summarized as follows:

Cladding cracks observed in all three specimens were intergranular it; nature, initiating on the outer surface.

In all cases, the cracks were associated with a fuel chip wedged in the fuel-cladding gap.

+

The cracks were branched and had the typical appearance of intergranular stress corrosion cracks.

In some cases, the cracks contained corrosion products.

The crack tips appeared blunt in some cases.

The cladding cracks did not appear to be associated with fuel pellet cracks.

The cracks were also unrelated to the seam weld in the cladding.

5.0 PRELIMINARY EVALUATION OF THE RESULTS AND CONCLUSIONS i

I Over the years that stainless steel cladding has been used in commercial reactors, there have been very few fuel rod failures. A majority of the instances of fuel rod failure involving stainless steel cladding were in boiling water reactor environments.' The failures observed were in the form of -

intergranular cracking (longitudinal and circumferential) starting at the outside in the high heat flux -

region. The cracks appeared to be of a stress-corrosion type, the source of stress being fuel-cladding interaction. The intergranular cracking of unsensitized Type-304 stainless steel cladding in the high-l purity water reactor environment was unexpected. However, on the basis of a detailed evaluation

(-

of failure in VBWR, Duncan concluded that commercial, unsensitized Type-304 stainless steel clad-l ding can become susceptible to grain boundary attack by l

t

Segregation of impurity elements at grain boundaries during reactor operation

Neutron irradiation induced defects (in the cladding), which enhance grain-:

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boundary susceptibility by migrating to the grain boundaries along with the '

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s 41

Accumulation of harmful species such as chloride ions at the cladding water interface.(4)

In the case of pressurized water reactor (PWR) environments (such as the Connecticut Yankee reactor), Type-304 stainless steel cladding had performed extremely well, in fact, there have only been three documented cases of fuel rod failures under PWR conditions. These were failures in the PM-3A core (5) (portable reactor in Alaska), the APPR core from an Army reactor (6) and in high burnup rodsin the core of the PR-3 Vulcain reactor (7). Failures observed in the Connecticut Yankee reactor are not directly comparable to the PM 3A or APPR observations in that the design of fuel rods and operating parameters were quite different. The PM-3A core consisted of annular fuel tubes made form 30-mil thick, UO -stainless steeldispersion clad with 7-mil-thick type 347 stainless steel.

2 The cladding was bonded to the fuel. Peak burnup in the fuel was approximately 50,000 MWD /

MTU The APPR fuel was of plate geometry with UO and 8 C in a stainless steel matrix, roll 2

4 bonded with Type-304 stainless steel cladding. (Cracks observed in these cases were intergranular in nature initiating on the outer surface.)

in the case of the Vulcain's core in the BR-3 reactor, the fuel rod failures were very similar to the Connecticut Yankee fuel rod cracks. However, the rods (containing UO2 pellets in cold-worked Type-304 stainless steel) were irradiated at significantly higher ratings to burnups > 40,000 MWD /

MTU. Peak heat ratings in theVulcaincore's fuel rods were approximately 11 kw/ft compared to the peak heat rating of 7 to 8 kw/ft in the batch 8 Connectucut Yankee fuel rods. The failures ob-served in Connecticut Yankee batch 8 fuel are unique in that they occurred under normal com-mercial PWR operating conditions.

Examination of the data presented in the previous sections of this report and in references (4),

(5), (6), and (7) indicates that the breaches in stainless steel cladding can occur by intergranular crack-ing. In addition, the results also suggest the cracks are stress-corrosion induced,with the source of stress being fuel-clad interaction. In the case of Connecticut Yankee fuel rods, high localized cladding stresses appeared to have been generated by the presence of fuel chips wedged in the fuel-cladding gap. Obsentation of chips in association with the incipient cracks in all three samples from unfailed rods examined strongly suggests that the primary defects in the failed rods were caused by a similar mechanism involving fuel chips. This, however, is not conclusive and needs to be confirmed in the next phase of work.

It has been reportedIO) that fuel chipping had been a common problem in the loading of pre-vious batches of Connecticut Yankee fuel rods. This observation suggests that the presence of fuel chips by themselves did not cause fuel rod failures. It appears that the presence of fuel chips com-bined with closure of the fuel-cladding gap is required to produce cladding stresses high enough to cause cracking. Examination of archive fuel pellets from batches 7,8, and 9 suggests that batch 8 fuel pellets may be sl:gitt:y less densifying than the fuel from batches 7 and 9 (based on pore size distribution). This may explain the significant number of fuel rod failures in batch 8 fuel. The proposed follow-on work described below is intended to provide evidence that could more con-clusively support such an explanation.

6 42 With the limited data available the following preliminary conclusions are drawn.

(1) Failure of stainless steel cladding can occur by intergranular stress-corrosion cracking from the outer surface in a PWR environment under conditions that produce highly i

localized stresses.

(2) Fuel chips wedged into the pellet-cladding gap appear to have produced high enough local cladding stresses to cause cracking.

(3) Primary cladding defects in the failed fuel rods may have originated by the same mechanism as in the unfailed fuel rods, but additional data are required to verify this.

(4) Cladding cracks appear unrelated to the presence of the seam weld in the welded and drawn starting material.

H 6.0 PROPOSED FOLLOW-ON WORK Based on an evaluation of data obtained thus far, work is planned in the following areas to verify preliminary conclusior" and to provide additional fuel performance data.

1 Obtain data to relate incipient cracks to long splits observed in failed fuel rods. This is to be obta.ined oy destructive examination c f a failed fuel rod at the location of the tight crack. Planned work includes metallogrsphy and scanning electron microscopy of the fracture surface. Attempts will be made to gather more evidence relating to the presence of fuel chips as a contributor tc the failure mechanism.

2. Additionally characterize incipient cracks by axamination at higher magnifications.

using the scanning electron microscope.

3.

Destructively examine rod 595-A10 from the batch 7 assembly G11 to determine if incipient cracks are present. Metallographic examination of the specimen at the location of eddy current indications will be carried out.

4.

Perform fuel density measurements on samples from both batch 7 and batch 8

. fuel rods to compare densification/ swelling behavior. A longitudinal metallog-raphic sample from the batch 7 fuel rod will be examined to verify dish closure.

5. Attempt to determine the chemical species involved in the stress corrosion cracking mechanism. Examination of an incipient crack location using Auger electron spectroscopy or ion microprobe analysis is planned.

Data obtained from this work will be useful in gaining better understanding of Connecticut Yankee fuel failure mechanisms.

43

7.0 REFERENCES

1.

Letter reports from Connecticut Yankee Atomic Power Company to USNRC-CYH 79-093 February 28,1979 and CYH 79-188 July 31,1979, Docket No. 50-213.

2.

M. Pitek, Northeast Utilities Services Company, private communication.

3.

Babcock and Wilcox letter report," Characterization of Connecticut Yankee Archive Fuel Pellets ', LR:79:5016-01:1.

4.

R. N. Duncan, " Stainless Steel Failure Investigation Program", final summary report: EUR AEC-GEAP 5530, February,1968.

5.

J. B. Brown, Jr., V. W. Storhok, and J. E. Gates, "Postirradiation Examination of the PM-3A Type 1 Serial 2 Core", ANS Transactions,1967, p 668-669.

6.

L. D. Schaffer," Army Reactors Program Progress Report", ORNL 3231, January,1%2.

7. J. Storre, "High Burnup trradiation Experience in Vulcan", Nuclear Engineering /nternational, February,1970, p 93-99.

8.

M. Pitek, Northeast Utilities Sciences Company, Private communication.

s

ML19341A968

Contents

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1.0 INTRODUCTION

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1.0 INTRODUCTION

2.0 BACKGROUND

SUMMARY

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