ML19259B964
| ML19259B964 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 05/24/1979 |
| From: | Groce R YANKEE ATOMIC ELECTRIC CO. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| WYR-79-60, NUDOCS 7906050276 | |
| Download: ML19259B964 (20) | |
Text
'
Telephone bl7 366-9011 Tw1 710-390-0739 YAllhEE ATOMIC ELECTRIC COMPANY a3 1
WYR 79-60
,Y_ T?1 20 Turnpike Road Westborough, Massachusorts 01581 NKEE
~~
May 24, 1979 United States Nuclear Regulatory Commission Washington, D. C.
20555 Attention: Office of Nuclear Reactor Regulation Mr. Dennis L. Ziemann Operating Reactors Branch #2 Division of Operating Reactors
Reference:
(a) License No. DPR-3 (Docket No. 50-29)
(b) YAEC letter to USSRC dated April 10, 1979 (WYR 79-47)
Dear Sir:
Subject:
Yankee Rowe Core XIII Fuel Examination Our letter Reference (b), transmitted for your use and infcrmation, a copy of the Exxon Nuclear Company report of the Core XIII fuel examinations conducted at Yankee Rowe in November 1978.
In a recent telephone conversation, your staff requested additional information relative to the examinations which was not included in this report. As a result of this conversation, we have enclosed a copy of a letter forwarded to us by Exxon, which sumcarizes in greater detail, the scope of the examination program and the observations noted. Uc are confident that this material will be of value to you.
This cr losure is not considered proprietary by the Exxon Nuclear Company and therefore, need not be withheld fron public disclosure.
Should you have any questions, please contact us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY Robert H. Croce Licensing Engineer RTT/cew Enclosure 79060502%
2283 276
Eff0N NUCLEAR COMPANY,Inc.
2101 Horn Rapids Road P. O. Box 130, Richland, Washington 99352 Phone: (509) 943-8100 Telex: 32-G353 KflW-79-01 R E C E l V c. D r
January 4, 19 JM 11 Core M
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Mr. Robert T. Chin Yankee Atomic Electric Company 20 Turnpike Road Westboro, MA 01501
Subject:
Fuel Examination of Discharged EllC Fuel
Reference:
Letter from R. J. Ehlers to Robert T. Chin, 11/22/78
Dear Bob:
The referenced letter described the results of an Exxon fluclear Company (Ef!C) fuel examination at the Yankee Rowe reactor.
This examination identified through-wall cladding perforations in two discharged assem-blies and raised the possibility that the EllC fuel design might be susceptible to fretting-induced fuel rod failures.
As a result of the first fuel examination, a second trip was scheduled to establish whether such a fretting-susceptibility risk truly existed.
Many more assemblies were to be examined in much closer detail, con-centrating on the area of the spacer offset corner.
This offset corner, which is unique to the Yankee Ro'.ze fuel design, occurs on two sides of
' ach assembly and was the location of the gross fretting damage observed e
on two assemblies during the initial examination.
The purpose of this letter is to report the results of this second fuel examination and to suggest a possible mechanism for the observed fretting failures.
SUMMARY
The source of the planum spring discovered in 1977 and the relatively high I-131 activity cbserved during Cycle 13 at the Yankee Rove reactor have bcon traced to through-vall fretting damage of tuo peripheral fuel rods in cach of tuo EllC fuct acocmblics.
Thecc fretting failurcs are thought to have occurred during the firov operating cycle (Cycle 12) of thace acccmblica uhen they were located at the periphcry of the rcactor core.
The cause of frctting cannot be established with absolute certainty.
It may be the result of loading damage when the shroudican acccmblica care first incerted into the corc periphery pocition.
The possibility that the fretting AN AFFILIATE OF EXXON CORPORATION
Mr. Robert T. Chin Page 2 January 4, 1978 occurred by crocc-flow currento from a damagcd or defcativa core baffic io alco cucpcsted based on the similarity of thcoc failurce with frctting failurco reported for three othcr operating RG'c.
A third accanbly judged to have failed by cipping tech-niques c:hibited no vicible cuidance of frctting damage even though it too had been located on the corc periphcry.
A fcurth anocmbly, which alco c:hibited a relatively high cipping cignal, io cucpcotcd to Lc contaminated with ficcion producto frcm a failed adjacent acccmbly.
There vac no indication of frctting damage on any of 9 othcr fuct acocmblice cramined, therchy indicating that the basic fuct acccmbly dccign ic free of any intrinsic cuccepitibility to fretting.
All 40 of the fuct acocmblica in the initial EllC reload, including the too visibly failed accc=blica, achieved burnup levelo ranging from 21,C00 to 32,200 N:C/NTU at the time of cchcduled diccharge.
Additional inci ht into the cauco of failure must avait t
rectenption of reactor operation at EOC 14 and a EOC li cramination of the core baffic and of acccmblics dio-charged from the core.
A.
SCOPE OF THE SECOND FUEL EXANIrlATI0tl A total of 13 fuel assemblies were visually examined during the period from flovember 29 to December 6,1978, at the Yankee Rowe site.
The upper tie plate was removed from 3 of these 13 assem-blies to allow partial withdrawal of individual fuel rods.
The core location of the 13 assemblies in operating Cycle 12 is depicted in Figure 1.
Selection of the 13 fuel assemblies was based on several factors, including core location, sipping ac-tivity, and assembly type (A or B).
Visual examination of the fuel assemblies was performed by under-water television techniques.
An underwater TV camera was suspended from a supporting bracket adjacent to the fuel assembly elevator.
The TV camera could be freely positioned in the horizontal plane; vertical movement was accomplished by raising or lowering the fuel assembly elevator.
The TV picture was monitored with equipment located on the fuel pool deck.
When appropriate, a video tape record of the TV picture was made.
Selected views of the fuel assembly were also recorded by taking Polaroid photographs of the TV monitor.
2283 278
Mr. Robert T. Chin Page 3 January 4,1979 The standard examination technique durng this second trip was to vertically scan the full length of the assembly at each of the two offset corner locations.
A horizontal scan was then obtained on each side of the assembly at the topmost spacer location.
One scan would be made on the upper part of the topmost spacer followed by a similar scan on the lower part of that spacer.
The assembly would then be rotated in 90 increments and similar scans obtained on each of the other three sides.
If the assembly was known to have a low sipping signal and showed no evidence of extensive fuel rod wear just above the topmost spacer, the visual examination was typically terminated at this point.
In several cases, the TV examination was continued on down the four sides of lower spacers on the assembly.
lhe upper tie plate of fuel assemblies A493, 8482 and 8502 were removed to expose the fuel rod upper end caps.
This step allowed individual fuel rods to be partially withdrawn from the assembly and rotated for a full view of regions normally obscured by the spacers.
B.
MAJOR OBSERVATIONS _
Of the 13 assemblies examined,11 showed no evidence of significant fretting type wear.
Due to dif ferential thermal expansion between the zircaloy fuel rod cladding and the stainless steel guide bars, the areas where the peripheral fuel rods had been contacted by the upper dimples of the topmost spacer were exposed to view (above the spacer).
In the 11 defect-free assemblies, there was typically only a shallow superficial scratch mark at these contact areas.
Although the depth of these contact areas was not measured, it was estimated that they were on the order of l-2 mils (0.001 + 0.002 inch).
Assemblies A493 and B482 had visible evidence of fuel rod failure at the upper two spacer locations on one side of the assembly. In each case, the defective fuel rods were located at or adjacent to an offset corner of the spacer.
(See Figures 2 and 3 for assembly maps of Type A and Type B assemblics, respectively.)
1.
Assembly A492 - Rod 8P Assembly A493 exhibited the greatest evidence of failure. The fuel rod in the offset corner position (location 8P) had fractured completely ( ff just above the bottom of the topmost The bottom of the offset corner of the topmost spacer spacer.
was also torn off (see Figure 4). The missing upper part of 2283 279
Mr. Robert T. Chin Page 4 January 4,1978 this rod was probably the source of the fuel rod plenum spring discovered during the 1977 reactor outage.
hd 8P exhibited extensive fretting wear at all other spacer locations.
At the second spacer from the top the cladding was worn completely through over an approximately 2" leng*F and the rod was on the verge of fracturing off again at that location (see Figure 5). The upper and lower dimple marks on the second spacer were worn away.
There was a through-wall fretting mark at the third spacer from the top, and deep fretting marks were apparent at the fourth, fifth, and sixth spacer locations.
The shoulder of fuel rod 8P, just above the lower end cap, was initially resting directly on the lower tie plate as though the lower end cap was missing.
During this examination, fuel rod 8P was pushed up by inserting a push-rod through a ficw hole in the lower tie plate.
When this was done, the lower end cap (still attached to the fuel rod) was exposed to view.
The end cap was badly worn on one side, although the rod serial number was still cicarly visible on the other side (see Figure 6).
Due to the fuel rod vibration and high downward forces on the fuel rod caused by differential thermal expan-sion, the lower end cap had apparently been driven into a flow hole in the lower tie plate.
Although the initial diameter of the flow hole was smaller than the initial end cap diameter, the wear-induced decrease in end cap diameter and increase in flow hole diameter apparently allowed the downward movement of the fuel rod.
2.
Assembly A493 - Rods 7P, 6P, and 80 Fuel rods adjacent to rod 8P also showed evidence of fretting damage.
Peripheral rod 7P had through-wall penetrations at the spacer spring and lower dimple locations of the topmost spacer (see Figure 7).
Deep fretting marks, typically 1/3 to 2/3 through the wall thickness were also observed at the upper dimple locations of the top three spacers of rod 7P and the adjacent peripheral rod 6P.
In most cases, a deep fretting mark was associated with a badly worn spacer dimple.
Rod 80. adjacent to rod 8P but an interior rod, showed exten-sive cladding damage to the topmost spacer location. This damage appeared to have been induced by the extensive vibration from adjacent rod 8P, after the upper end cap fractured off.
2283 280
Mr. Robert T. Chin Page 5 January 4, 1979 1
3.
Assemaly B482 i
Fretting-type damage was also observed on two adjacent rods in assembly B482.
The fretting occurred at dimple and spring locations on the top two spacers.
Through-wall penetrations occurred at locations corresponding to the bottom dimples of the topmost spacer and the top dimple of the second-from-the-l top spacer.
Again, through-wall penetrations were accompanied by severely worn spacer dimples (see Figure 8).
There was no evidence of fretting-type wear on any of the other fuel rods in this assembly, including fuel rods at the other offset corner location.
4.
Assembly B502 Assembly B502 was closely examined because, like assemblies A493 and B482, it exhibited a relatively high sipping signal.
The upper tie plate of the assembly was removed to allow close examination of both peripheral and interior foci rods.
Although only approximately 10% of the fuel rods was closely examined, there was no indication whatsoever of any significant fretting-type wear.
To detemine the exact location of any failed rods in this assembly would require a much more exhaustive examination effort and would probably utilize eddy current techniques for detecting very small cladding failures.
No additional exami-nations by ENC are planned at this time.
5.
Assembly B490 Assembly B490 initially exhibited a relatively high sipping signal and was judged to have failed.
On a resip of assemblies A493, 8482, 8502 and B490, however, assembly B490 exhibited a much lower I-131 sipping signal than the visibly failed assembl ies.
A summary of the two sipping results is shown below:
I-131 Relative Sip Signal Assembly _
Initial Sip _
Resip A493*
124 132 B482*
21 27 284 22 B502 83 8
B490
- Visibly failed 2283 281
Mr. Robert T. Chin Page 6 January 4, 1979 The first three assemblies (A493, B482, and B502) were all located t,n the core periphery during Cycle 12.
Assembly B490 was never on the core periphery but did share a common face with badly failed assembly A493 during the second operating cycle, Cycle 13.
Al though membly B490 was not disassembled during the visual examination, there was no evidence of fretting wear of any of the peripheral fuel rods.
Due to the foregoing circumstances, it is tentatively assumed that assembly B490 may not have actually failed but may have exhibited a higher-than-normal sipping signal due to cross-contamination of fission products from adjacent assembly A493, which contained the severed fuel rod.
C.
DISCUSSI0ft Mechanical damage appears to have occurred to fuel assembly A493 during loading.
A vertical streak was observed along the spacers on the side where the fuel rods had failed.
In some video tape views, this streak looks like an extended scrape mark.
Mechanical damage during loading would have the effect of flattening out the dimples in the peripheral strap and causing fretting vibration.
fle definitive evidence of mechanical damage was observed on assembly B482.
The observed failure character istics in assembly A493 and B482 are very similar to published accounts of fretting failures in fuel assemblies located on the core periphery of two commercial PWRs, Zorita 1 and Point Beach 1 In both of these cases, severe fretting and fuel rod failure occurred due to flow-induced vibration of peripheral rods of peripheral assemblies.
The cause of the flow-induced vibration was eventually traced to leaks in the core baffle, which allowed a concentrated coolant cross-flow to impinge on the sides of operating fuel rods.
Published accounts of these fretting type failures are reproduced in the Appendices A, B, and C.
During a 1978 outage at a European PWR, similar fretting failures of peripheral fuel assemblies were obs'erved.
Close inspection there also showed that a gap had developed between two adjacent sections of the core baffle.
The utility staff has tentatively concluded that the fretting failures were caused by vibrations induced by flow leakage through the baffle.
2283 282
Mr. Robert T. Chin Page 7 January 4, 1979 Based on these three known cases of fretting failures in peri-pherally-located fuel assemblies, it i possible that a similar mechanism might be responsible for the observed fretting in Yankee Rowe assemblies A493 and B482.
Supporting this hypothesis is the fact that the fretting damage occurred only on peripheral rods.
The fact that the fretting damage was greatest near the top of the assembly would be understandable if a gap in the core baffle had developed at that axial location.
One inconsistency in proposing this baffle gap theory is the lack of observed fretting damage on peripheral rods of assembly B502, which exhibited a relatively high sipping signal.
It may be that failures were present but undetectable without withdrawing more rods.
The fact that assembla 8502 was located on the core periphery during Cycle 12, just like assemblies A493 and B482, increases the probability that it too has failed.
If baffle gaps are indeed the cause of the observed fretting damage, it should follow that simar failures will continue te The fact that prior fa lures were not observed is under-s occur.
standable because earlier fuel was enclosed in a protective assembly shroud.
During Cycle 13, however, new ENC assemblies were located in the same core positions as the failed assemblies had been in Cycle 12.
Since the fuel design was unchanged, the Cycle 13 assemblies should also have been exposed to the cross-flow of potential baffle gaps. These assemblies had not been sipped.
At the time of this writing, the Yankee Rowe reactor had just achieved full power with very low I-131 activity reported.
Con-tinued operation will be required to assess whether additional fuel failures may occur.
D.
CONCLUSIONS ENC is reviewing the new all zirc fuel assembly design to strengthen its resistance to the fretting characteristics observed on assemblies A493 and B482.
We would like to monitor the I-131 and I-133 activity in the primary coolant during the current Cycle 14 and would appreciate receiving the daily values normally collected.
If the I-131 coolant activity should become significant, the following actions should be planned for the next scheduled outage:
1)
Sip assemblies which had occupied core positions A7 and and 83 during either Cycle 13 or Cycle 14. At that time we will conduct a visual examination on those four assemblies and any others with a high sipping signal.
2283 283
Mr. Robert T. Chin Page 8 January 4, 1978 2)
Examine the core baffle at locations opposite fuel assembly position A7 and B3.
It may be worthwhile for you to contact the Point Beach 1 technical staff to learn more of their experience with core baffle gap detection and repair.
I hope the information transmitted in this letter report is of interest to you and your colleagues.
We are ser. ding, under separate cover, complete copies of all video tapes recorded during both exams.
Sincerely yours,
.AU K. N. Woods, Manager Fuel Performance KNW:wrg cc: RJ Ehicrs File /LB 2283 284 i
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FIGURE I Thirteen Assemblies Visually Examined During Second EllC Fuel Examination at Yankee Rowe, December 1973.
Core Locations Refer To Cycle 12 Relatively liigh Sipping Signal 2283 285
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.t Severed Fuel Rod 8P at Top Spacer location, l
I uel Assembly A493
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4 Exposed Fuel at Second-From-The-Top Spacer g
1 Location of Rod 8P, Fuel Assembly A493 F
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-A 2283 288 e
i Figure 6 Deformed Lower End Caps of Rod 8P (left) and Rod 7P (right), Assembly A493
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Figure 7 Through-wall Penetration of Rod 7P At Lower Dimple Locations of Top Spacer, Assembly A493
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APPENDIX A GAP IN BAFFLE - ROD VIBRATION & DAMAGE Zorita 1 - 1971, 72 During their first refueling (1971), VT disclosed 2 outermost broken fuel rods and one partly damaged rod in a peripheral assembly.
They were broken at the 2nd grid location from the top and the damage was confined to one corner of the assembly; no evidence of loss of cladding integrity was found in the remainder of the core.
Initially, it was r otulated this damage was the result of a handling incident prior to operation.
The plant was refueled and operated through the 2nd cycle.
Similar observations were made at the end of the 2nd cycle (1972) in the same core position and in the nirror-image position on the other side of the reactor. The cause of the damage was then identified as high-velocity-coolant-cross-flow leaking through gaps in the corner joints in the core baffle. The cross flow jetted across a slight gap in the baffle (20 to 40 mils wide) and caused excessive rod vibration and eventual fretting through the cladding in the grid support areas.
Corrective action was taken during the 1972 refueling; the baffle joints were repaired (they peened the gap closed) to eliminate the leakage.
The absence of damage during the cycle 3 refueling inspection confirmed the efficacy of this reoair.
Baffle designs in later reactors eliminated the gaps completely.
(yz,bcm)
Visual Techniques 2283 291 A-1
APPENDIX B WATER IMPINGEMENT FRETTING, RAPID POWER ESCALAfION - FUEL DAMAGE Pt. Beach 1 - Dec 75 - refueling shutdown During refueling, the Core Loading Supervisor noted something protruding from the side of FA D-03 as it was being lowered into the FA*
upender.
Movement was stopped and an underwater TV inspection showed that a section of fuel rod was missing.
The damaged FA was then moved to the fuel inspection periscope in the spent fuel pit where a more detailed exam was made.
The core location K-6 from which FA D-03 was removed was then inspected with no abnormalities noted.
FA's in the area of K-6 were removed from the core to facilitate inspection of these FA's and the core support plate.
An 11 in. section of fuel rod was found and removed by an underwater vacuum cleaner.
Inspection of the FA's adjacent to D-03 revealed several fuel fragments wedged between fuel rods and the grids on FA D-34.
This FA also had several scratches adjacent to the area of major damage in FA D-03.
D-34 was later removed from the core as previously scheduled in the refueling pro-cedure.
A small plug like object was removed from the top of the bottom nozzle of FA F-07.
The detailed visual exam of D-03 indicated considerable damage to fuel rods No.12 and 13.
These rods showed displacement verti-cally and horizontally.
Sections of these rods were missing.
The cladding was worn, gouged, cracked and sections were missing.
The grid flow vanes were worn and bent.
Fragments of fuel and clad were found in adjacent areas of the FA.g} }g]
- Fuel Assembly B-1
D-03 was fabricated by ((. It was of a prepressurized low parasitic design with fuel rods off the bottom nozzle. The initial inspection of the FA showed no abnormalities. D-03 was initially loaded into core position H-1 for Cycle 2 and was relocated in K-6 for Cycle 3. No abnormalities were noted in the FA during Cycle 2 - 3 refueling operations. (( stated that the fuel failure was probably due to water impinge-ment through-the baffle plate while the FA was in position H-1. Previous experience in 2 foreign plants had indicated that flow through the stitch weld joining in the baffle plate or bolted baffle plates had occurred in core locations similar to H-1. This condition can result in a water impingement on the corner or near corner fuel rods of the assembly at this position, thus leading to vibration and fretting wear of 1 or more fuel rods. It was postulated that this wearing action eventually produced pin-holes through the clad and the fuel rods became water logged during refueling shutdown. The initial escalation to power for Cycle 3 was at a higher rate than the current operating guidelines permit. A sharp increase in coolant activity was noted between 40 and 50% power resulting in bursting of these damaged fuel rods due to internal steam pressure. The FA in core position H-1 for Cycle 3 was relocated in position J-3 for Cycle 4 operation and a new FA was loaded into position H-1. These 2 FA's were to be carefully inspected at the end of cycle 4. Core position J-3 has no control rod so that any possible damage to the FA in this position could not prevent control rod dropping action. Controlled power escalation for Cycle 4 was to follow W guidelines which allow for expelling of any water which may be in the FA due to B-2 2283 293 F
pin hole leaks, thus preventing bursting cf cladding due to internal steam pressure. Careful monitoring of the coolant acticity was to be followed to determine any future fuel failures. It was not known if any fuel pellets remained in the RCS from Cycle 3 operation or Cycle 4 reloading. All fuel fragments observed during the refueling inspection were removed. Fuel pellets that may have dropped through the lower core support plate during refueling could possibly cause local flow blockage but this was considered tolerable in this axial core location. Fuel pellets retaining their integrity would not generate fission products since they are not exposed to core nuetron flux being captured below the bottom grid plates of the FA's. If the pellets disintegrated, it was expected that temporary minor coolant activity increases would be detected. The chemical and volume control system purification demineralizers and filters would remove particulate and dissolved materials from this source. Coolant activity levels were not expected to exceed those experienced in Cycle 3. It was planned that all fuel and the lower core barrel would be removed during the Cycle 4-5 refueling for ISI. This would permit a thorough fuel inspection and investigation of baffle plate joints to further evaluate the cause of the failed fuel rods. (cli,dcc) B-3
APPEllDIX C FUEL CLADDIIG DAMAGED - WATER IMPIrlGEMEE THROUGH PAFFLE PLATE Pt. Beach 1 - Dec 76 During the inspection of fuel discharged from the core following the refueling outage of Oct-flov 76, the presence of through-wall cladding defects in one rod of FA G07 was confirmed after physically raising the rod to expose the area previously hidden by the fuel grid straps. Previous VT's on 1 and 3 t!ov 76 disclosed grooves in the cladding, but not through-wall defects; however, the rod was seen to have moved down s 3/4 in. from its normal position in the grid straps, the bottom of the rod coming to rest on the bottom grid plate. Fuel sipping of the FA on 4 and 17 fiov 76 did not detect cladding degradation, probably due to the low burnup condition of the G07 FA. A similar fuel cladding problem was discovered on 12 Dec 75 and reported in I.28. The H FA G07 had pre-pressurized fuel rods with zircaloy cladding and with one fuel rod having through-wall cladding defects which were caused by the rod vibrating against the rod retaining grid springs. Vibration of this particular fuel rod was related to the crossflow of water through the baffle plate, as described in I.28. Peening of the core baffle plate joints was completed in flov 76 for this unit and planned for Unit 2 in Mar 77 to prevent a recurrence of this event. The G-07 FA had been in the core only one cycle and they were planning to reinstall the assembly in a future cycle provided the damaged rod could be removed and f1RR approved operation with one rod They planned to mill a slot directly below the No.12 rod in missing. the bottom nozzle assembly through which the rod could be removed. (efk) 2283 295 C-
T mHQ ./ UNITED ST AT ES 4 / , t NUCLEAR REGULATORY COMMISSION f REGION til g.g g g,gl f 799 ROOST V E L T HC AD g G L E N I L L Y N. IL LINOf S 60137 o.....- APR 2 51979 T. ' e;t No. 50-346 Toledo Edison Company ATTN: Mr. W. A. Johnson Senior lice President Operations Edison Plaza 300 Madison Avenue Toledo, OH 43652 Gentlemen: This refers to the investigation conducted by Messrs. J. E. Foster and J. E. Kohler of this office on January 6, 14 and 29, 1979, of activities at the Davis-Besse site, authorized by NRC Operating License No. NPF-3 and to the discussion of our findings with Messrs. Domeck and Novack of your staff at the conclusion of the investigation. This investigation was conducted due to concerns expressed by one of our inspectors relative to operations, evaluation of system responses, and facility changes at the Davis-Besse facility. The enclosed copy of our investigation report identifies the areas examined during the investigation. Within these areas, the inves-tigation consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. No items of noncompliance were identified during this investigation. However, we continue to be concerned with certain aspects of your management control system involving facility changes. Specifically, failure to control a facility change and inadequate communication with the Office of Nuclear Reactor Regulation resulted in issuance of a change to your Technical Specifications prior to completion of the engineering necessary to implement the change. The example of taking timely actions on facility change requests contained in not this report, a matter of concern, was discussed in our August 1978 meeting with you. This matter was again discussed with you during our meeting on April 18, 1979. The details of the April 18 meeting will be documented in a separate report. 228l3 296 4 7 90623 0 l l6i1
a APR 2 51979 Toledo Edison Company In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room, except as follows. If this report contains informatior. that you or your contractors believe to be proprietary, you must apply in writing to this office, within twenty days of your receipt of this letter, to withhold such information from public disclosure. The application must include a full statement of the reasons for which the information is con-sidered proprietary, and should be prepared so that proprietary information identified in the application is contained in an enclosure to the application. We will gladly discuss any questions you have concerning this inspection. Sincerely, Q. r,..,.h,h.: q tt: ~ ' James G. Keppler Director
Enclosure:
IE Investigation Report No. 50-346/79-06 cc w/ enc 1: Mr. T. D. Murray, Station Superintendent Central Files Reproduction Unit NRC 20b PDR Local PDR NSIC 2283 297 "c Harold W. Kohn, Power Siting Commission}}