IR 05000309/1990008
| ML20056A943 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 07/24/1990 |
| From: | Terao D, Winters R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20056A941 | List: |
| References | |
| 50-309-90-08, 50-309-90-8, NUDOCS 9008100187 | |
| Download: ML20056A943 (11) | |
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S._ NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-309/90-08
. Docket No. 50-309 License No. DPR-36-
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. Licensee:- Maine' Yankee Atomic Power Company 83 Edison Drive 4.
Augusta, Maine 04336 A
Facility Name: Maine Yankee Atomic Power-Station C
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Inspection At: Wiscasset, Maine p
Inspection Conducted: May 7-11, 1990 n?
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. Inspectors:
1 E@ *) 6 R. W. Winters, Reictor Engineer, MPS, EB,
'date DRS, Region I Approved-by:
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D. Terao, Chief, Materials & Processes date
~Section,-Engineering Branch,-DRS,.RI
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Inspection Summary:
A' routine unannounced inspection.was conducted from May-7-11, = 1990 (Report No. 50-309/90-08) of'the licensee's inservice inspection program including outage performance and the control'of.the program.
The-
- areas inspected included eddy. current' examination of-steam generator tubes,
-itemsLexamined during the 1990 refueling outage and comparison 'of the work-
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performed within the 10 year inservice inspection program.
s Results: During eddy current. examination of the~ steam generators significant cracking was found in-the tubes at the top of the tubesheet. The licensee expended significant_ effort to assure all cracked tubes were identified and removed from service. No' violations or deviations were identi fied.
However,
- itlwas noted that the licensee's program for monitoring primary-to-secondary-system leakage could be improved, particularly considering the potential for continued circumferential cracking of the steam generator tubes.
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9008100187 9009o3
{DR ADOCK 050003o9 PDC
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.p DETAILS 1.0 Persons Contacted
Maine Yankee Atomic Power Company
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- R. Bichail, Outage Manager
- R. Blackmore.. Plant Manager
- A. Cayia,. Manager, Operations C, Eames, Nuclear' Engineering
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- T. Gifford, Project Engineering Section Head J. Hebert, Engineering Manager
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P. Melhorn, Inservice Inspection Engineer
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- S. Nichols,. Licensing Section Head P. Plante,-Nuclear Engineering
- P. Radski, Chemistry Section Head
- C, Shaw, PED Section. Head
~*L. Speed, Engineering Supervisor Yankee Atomic Energ L n oration C
C.~Larson, NDE Engineert
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State of Maine
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- P. Dostie, NucleariSafety Inspector
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Combustion ~ Engineering Corporation
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R. Maurer, Nondestructive Examination Supervisor i
Conam Nondestructive Examination Company
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M. Chambers, Nondestructive Examination Supervisor
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Ebasco Services Corporation
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R. Zeber, Nondestru'ctive Examination Supervisor United States Nuclear Regulatory Commission
- R.' Freudenberger, Resident Inspector
- C. Holden, Senior l Resident Inspector
' E. Leede,. Project Manager i
- R. Wessman, Project Director, PDI-3
- Denotes thosca attending the exit meeting.
The inspector'also contacted other administrative and technical personnel during the inspection.
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2$0-Introducticn
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Maine Yankee -- a three-loop pressurized water reactor plant -- was licensed
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for commercial' operation on September 15, 1972. ~Since commercial operation,
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Maine Yankee has been sensitive to water leaks into the condensers and has repaired leaks of very small magnitude.
The condensers were retubed from t i aluminum-brass to copper-nickel in 1976, from copper-nickel to stainless steel in 1979, and then from stainless steel to titanium in 1985.
Since i
the last retubing, the condensers have remained leak-tight.
During the 1990 outage, the last remaining copper.in the secondary system (feedwater y
heaters) 'Is being replaced with stainless steel.
In an effort to 2 top
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i tube denting in the steam generators, the licensee initiated boric acid
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treatment of the. secondary water in 1981.
This treatment was' successful.
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in arresting the denting.
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The lack of significant degradation in the steam generator tubes prior to
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S-the present outage reflected the adequacy of the : licensee's program to.
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control secondary water chemistry.
Each steam generator wa manufactured
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with 5703 Inconel= tubes..The history of tube plugging is shown in Table 1.
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TUBE PLUGGING HISTORY j
i YEAR Steam Generator
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During Manufacture
1978 (Note 1)
5
1982
1985~
22
1987-
j 1988
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1 Note 1-- Tubes plugged as a result of a rim cut due to denting.
l 3.0 Inservice Inspection i
Program Review-l The facility is in the third period of the second 10 year inservice inspection interval that started on December 28, 1982 and will end on
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December 28, 1992.
The facility is currently committed to meet the
requirements of the 1980 Edition of the ASME Boiler and Pressure Vessel
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Code,Section XI up to and including the Winter 1980 Addenda.
The 1990 outage inservice inspection (ISI) plan and the 10 year program were selected for review, ine inspection was performed with respect to completion of the scheduled items o.9 to ascertain that tFc schedule agreed with the 10 year plan onJ ine current inspection period plan.
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The inspector determined that the required examinations ~were scheduled andLcompleted.': The 10 year program had been converted from a manual system to a computer based _ system during the past operating cycle. The new computer-based system was being used to_ maintain.the' program and track.its status to assure that all required examinations were completed
'within theLCode and regulatory mandated time frame.
Implementation q
The~ inspector reviewed the master controlling document describing the status of the ISI activities and interviewed the cognizant licensee and contractor personnel to determine if the program was adequately controlled.
The master a
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schedule was; established to provide positive status jndication at each
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'q step of,the testing and data review.
This schedule was updated daily and at.the^ time of the inspection was found current. The inspector noted that inspections deleted from the original schedule were clearly identified and would be rescheduled for the next ISI outage.
Additions were also clearly -
identified and the status tracked.
In all, there were less than 20 items
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.that,had'been~ deferred.
The number of deferrals is important since there j
is only'one outage left-in the interval to complete all required-inspections.
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During the inspection, no nondestructive examinations were in progress l
because all but a few had been completed.
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Conclusions The. implementation of the licensee's program for ISI has been enhanced by the use of a'_ computer-tracking system to assure tt.at all of the required-inspections are' performed..The_ method.of controll.ing work in orogress
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also~ provided a positive and effective means for determining inspections left to be performed.
From the number'of inspections that had been
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performed at the time of this inspection, it appears reasonable'that the licentee will be able to complete all required inspections for the interval
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on schedule.
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i 4.0 ' Steam Generator Inspection l
During the.latter part of the cycle, some radioactivity was noted in the
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steam-jet air ejector indicating ' radioactivity from steam generator Number 1.
- No radioactivity was noted for the other two steam generators.
During
-shutdown,' radioactivity was noted in the blowdown from steam generator
Number 1.
Although higher radioactivity was noted than had been identified
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in any of the steam generators previously, this radioactivity was slightly
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above the level of detection and therefore indicative of an extremely small
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leak.
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'Thel licensee developed an extensive program for the inspection and
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examination of the: steam generators.
This program included the:following:
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Leak test - With the secondary side of the steae generator filled to cover the tube bundle, the secondary side was pressurized to 160 psig.
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The licensee observed the primary side of the. tubesheet for leaks using' video cameras.
Both sides of all steam generators were scheduled for this test.
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Bobbin coil'- A full length inspection of 1543 tubes was scheduled
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in steam generator Number 3.
The bobbin coil Technic:1 5pecification examination was performed-using-a
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standard 0.580 inch diameter bobbin coil at a withdrawal speed of 24 inches per.second.
Frequencies ~were 25, 100, 400 and 600 kHz.
Motorized Rotatir,g Pincake Coil (MRPC) - An inspection was Lscheduled
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of the area 2 inches above to 2 inches below.the top of the tubesheet j
for all accessible' tubes on both the hot and cold leg sides of all three steam generat, ors.
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The MRPC primary inspection.was performed using a withdrawal speed of 0.2
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inches per second, at 200 rpm.
Frequencies were 50, 200, 300, and 400 kHz.
The diagnostic MRPC technir,ue was performed at a withdrawal speed nf 0.1
inches per second, at 2P0 rpm.
Frequencies were 50, 200, 300,' and 400 kHz.
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- Ultrasonic Examination - All circumferential indications found by i
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eddy current examination and other distorted indication signals as
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determined during th~e eddy current testing'was scheduled for ultrasonic en.mination.
Ultrasonic testing was done with a 45 rotating shear-q
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wave.
Axial.and circumferential scans were done.
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Results I
i After plant shutdown, a static leak test in steam generator Number 1 did
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not~show any leak.
However, af ter pressurizing the secondary side to.160.
j psi', one tube was found that dripped and one tube had a wet spot. A similar pressure test in steam generator Number 3 showed one tube with a leak that j
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allowed'one' drop per 3 minutes.
Steam generator Number 2 had not been j
tested at the time of this inspection but the licensee planned to inspect
this unit as soon as the equipment was available from the other steam o
generators.
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The eddy current examinations were still in progress at the time of the inspection.
However, the licensee reported the following results subsequent i
to the inspection:
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t Steam' Generator Number i
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As a result of the leak test, two leaking tubes were identified. One y
, (R82L113) of these tubes had.a drip every four minutes, the other (R82L33)
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The MRPC examination of 5682 tubes in the hot leg found 23 pluggable
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circumferential indications and 1 pluggeble pit No pluggabl>s indications i
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j Steam Generator Number 2
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No' leaking tubes were identified during the leet test. The MRPC examination I
F of 5655 tubes in the hot leg identified three plugeble circumferential indications and 1 pluggable pit.
No pluggable indications wert found in
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i Steam Generator Number 3 i
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The leak test identified one tube that had a drip every three minutes.
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This leak was^ confirmed by both bobbin and MRPC examination to be an axial i
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crack at the-second egg-crate support.
MRPC examination of 5687 tubes in the hot leg found the following pluggable indications:
the axial indication above, one distorted signal indication,
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i four circumferential cracks, and two pits, One circumferential crack was found by MRPC in the cold leg.
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The bobbin coil examination of 1543 tubes in this stsam generator found
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four pits greater than 40% through-wall approximately 6 inches above the
tubesheet,.18 pits between 20's and 39% througt vall, and 20 pits less than 19% through-wall.
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. Ultrasonic testing results were not available at ths time of the licensee's'
subsequent report. The final number of tubes to be plugged as a result of
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this steam generator inspection will be determined by the licensee af ter
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all of the results are evaluated.
Primary-to-Secondary Leak Detection
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The inspector discussed.the primary-to-secondary leak rate monitoring
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program with the licensee.. The present program meets the regulatory
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requirements and.in the past has been adequate for_ assuring the safe
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- operation of 'the plant as has bes n demonstrated by the integrity of the steam _ generator tubes. This program was set up to-determine the total
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leak' rate within the primary system and did not specifically address.
primary-to-secondary leakage.
In view of the steam generators exhibiting
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circumferential cracking at the top of the tube sheet, the licensee was
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considering improvements to this program that would assure steam generator leaks would be detected more quickly so that appropriate action could be taken.
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L Copper Replacement Program
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As a result of denting experienced in the stea:n generators, the licenseeL l
i initiated a program to replace the copper bearing tubes in the secondary
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ll water system heat exchangers.
This program was completed during the 1990
i refueling outage.
In all cases except for the main condensers, stainless I
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steel was ised to replace the copper bearing alloys. The main condensers
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have been retubed twice, the first time from tube of a copper bearing
alloy to. stainless steel and the second time from stainle u steel to
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titanium. Table 2 shows the schedule for this program.
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F TABLE 2 FEEDWATER C0pPER CHANGE OUT SCHEDULE
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' 'ponent Tube Surface Maximum Operating Year i
Com F_t2 Temperature 'F Replaced t
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Main Condensers 200,000 180 1978/1980 First Point i
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Feedwater Heaters 38,970 450 1980
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Moisture Separator
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Reheaters 66,400'
550 1984
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Main Condensers 200,000 180 1985 Second Point
. Feedwater Heaters 19,590
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' Sixth Point Feedwater Heaters 29,290 180 1988
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Fifth Point External i
Drain Coolers 7,450 190 1988 i
Gland Steam Condensers 790 210
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-Third Point Feedwater
- Heaters 20,380 320 1990 Fourth Point Feedwater Heaters.
24,380 270 1990
-Fifth Point Feedwater 23,590 220 1990
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Heaters
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l This program has. effectively arrested the denting previously experienced in k
the steam generators.
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l Conclusions
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The licensee recognized the implications of'the top of tubesheet cracking
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and is using the latest technology to identify the extent of the problem.
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This includes using a special eddy current probe developed by the L
contractor for identifying circumferential cracks at the top of the l
tubesheet and ultrasonic inspection to get accurate sizing of these cracks..
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i The licensee has established an acceptable method of data analysis that I
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tube by reducing the severity of an indication from defective.to degraded i
without concurrence of another certified individual,
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.The licensee has completed the program for replacing the copper bearing
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t alloys.in the secondary water system.. This program was designed to
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eliminate centing in the. steam generators at the support plates.
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current. inspection of the steam generators over the past few cycles have:
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5.0 Therm'alsnieldPositioningPins
Introdet*;n a
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The thermal shield in the reactor vessel at Maine Yankee has 17 positioning
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pins equally spaced around the bottom and 9 positioning pins around the
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top. The original design for locking these pins involved threading the
pins in place and tightening, placing a locking bar-in a slot milled in
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the pin for the purpose, and. welding this bar to the thermal shield in the i
counterbore'provided.
Inspection of the thermal shield positioning pirs i
=during the 1984 outage revealed that three of these pins on the top were i
dislodged and one was loosened. On the bottom, seven positioning pins were. loosened and one locking bar was missing.
Repairs were performed by
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, tightening the loose ~ pins and replacing the locking bars with hollow
.i spring-type lockpins inserted parallel to the axis of the positioning pin
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' Findings The licensee has designated the lower positioning pins as A-Q t'd tLe upper positioning pins as R-Z.
Table 3 indicates the results and corre uive
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action for each of these. pins during the 1984 and 1990 inspections. Note
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.that the 1990 results are preliminary since inspections were in progress
during the NRC observations.
TABLE 3
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INSPECTION FINDINGS FOR POSITIONING PINS
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i Pin A
1984.not reported 1990 not reported B
1984 acceptable 1990 acceptable
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.C 1984 Positioning pin replaced and loaded spring lockpin
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installed-1990 Lockpin missing, positioning pin rotated with
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approximately five threads showing
D 1984 not reported
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1990 not reported
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1984 not reported l
rE 1990 not reported i
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1984 Positioning pin tightened,.lockpin installed L
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1990 Lockpin missing, positioning pin rotated l
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G-1984 Positioning pin tightened, lockpin installed
1990 Lockpin missing, positioning pin rotated j
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H 1984 acceptable j
1990. acceptable
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1984 acceptable l
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1990 acceptable
J 1984 not reported
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1990 not reported
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K-1984 Positioning pin tightened, lockpin installed '
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1990 Lockpin missing, positioning pin rotated
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1984 Positioning pin tightened, lockpin installed
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1990 Lockpin missing, positioning pin rotated
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1984 not reported
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N 1984 Positioning pin tightened, lockpin installed
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1990 Lockpin backed 1/4 to 1/2 1nch out and damaged
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1984 not reported i
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1990 not reported 4i P
1984. Positioning pir tightened, lockpin installed D
1990 acceptable, lockpin in place Q
1934 Positia m g pin tightened,-lockpin installed i
1990 Lockpin missing, positioning pin rotated
R 1984 not measurable
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1990 not measurable
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1984 acceptable
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.n 1990 not reported l
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1984 acceptable
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1990 not reported l
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1984 acceptable
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1990 not reported l
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1984 Positioning pin replaced, lockpin installed
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1990 not reported
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1984Positioningpintightened,lockpininstalled
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g 1990 Lockpin backed 1/4 to 1/2 inch out of position
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X 1984 Positioning pin replaced, lockpin installed j-1990 new lockpin installed
Y 1984 Positioning pin replaced, lockpin installed
1990 new lockpin installed
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Z 1984 acceptable 1990 not reported The licensee performed an extensive search of the reactor' vessel to recover the missing lockpins from the reactor vessel.
From observations of the
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-condition of the lockpins recovered, the licensee determined it was unlikely
that any pieces of lockpins would be small enough to pass through the water
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flow holes in the fuelfsupport plate and enter the fuel bundles.
Although the failure analysis had not been completed at the time of this inspection, the initial;results indicated that fatigue of the 4XX-series stainless steel lockpins was the primary cause of failure.
A secondary'
cause could be attributed to primary water stress corrosion cracking in the hightstress portion 9f the spring lockpins.
To prevent recurrence of
lockpins working out of position or fracturing, the licensee has designed
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a.new solid lockpin using the same 3XX-series stainless steel as that used i
in thetpositioning pins. This design will alleviate loosening due to thermal expansion, and the solid design will prevent. fatigue fracture.
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Conclusions
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The licensee's inspection of the positioning pins in the thermal shield has been thorough.
The six loose lockpins and the two partially-F 'odged
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lockpins of the total of nine installed indicate that the former
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was deficient. The revised design developed by the licensee appeL. to
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l have addressad these deficiencies.
Furthermore, the licensee's preliminary conclusion indicated that if all of the missing pieces cannot be found, lthese loose parts are too large to enter the fuel bundles and can not cause
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damage to the fuel.
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e 6.0 Management Meeting Licensee management was informed of the scope and purpose of the inspection at the entrance interview on May 7, 1990. The findings of the inspection
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were discussed with licensee representatives during the course of the inspection and presented-to licensee management at the May 11, 1990 exit interview (see paragraph I for attendees).
At no time during the inspection was written material provided to the licensee by the inspector. The-licensee did not indicate that proprietary information was involved within the scope of this inspection.
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