ML20247K097
| ML20247K097 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/19/1989 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 89-348, NUDOCS 8906010209 | |
| Download: ML20247K097 (9) | |
Text
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4 VIRGINIA ELECTRIC AND POWER COMPANY RIcnMoND, VIRGINIA 23261 i
I May 19, 1989 I
U.S. Nuclear Regulatory Commission Serial No.89-348 j
Attn: Document Control Desk NO/JBUR2 4
Washington, D.C. 20555 Docket No. 50-338 License No. NPF-4 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA' POWER STATION UNIT 1 STEAM GENERATOR CATEGORY C-3 INSPECTION RESULTS REQUEST FOR STARTUP. APPROVAL North Anna Unit 1 experienced a reactor trip on February 25,1989. During the reactor trip event, a tube leak occurred in the 'C' Steam Generator. Because the unit had been in a power coastdown and based on the presumed scope of work to repair the steam generator tube leak, Virginia Electric and Power Company elected to begin the refueling outage previously scheduled to begin in April 1989. During this refueling outage, eddy current examinations of the steam generator tubes required by the augmented eddy current examination program described in the facility Technical Specifications were conducted. The results of these steam generator tube eddy current inspections have been classified as Category C-3 for each of the unit's three steam generators.
In accordance with the requirements of Technical Specification 4.4.5.5.c, prompt notification of the condition was made to the NRC pursuant to 10 CFR 50.72 and a Licensee Event Report (LER) No. 89-004-00 was submitted on May 5,19o9 for the category C-3 inspection results on each of the three steam generators on Unit 1.
The purpose of this letter is to request NRC approval to return to power operation following the current refueling outage presently scheduled to be completed about June 15,1989. In accordance with Technical Specification Table 4.4-2, NdC approval is required prior to operation when the inspection results of any two of the steam generators are classified as category C-3.
The attnched evaluation provides a description of the steam generator inspections and provides the basis for the determination that approval for resumption of power operation should be granted. The technical basis for the inspections and corrective actions VAll be provided in a separate letter by May 26,1989 as discussed with the NRC Project Manager fod t
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l i Should yon have any questions or require additional information, please contact us immediate%
- -Very truly yours, l
t-
. q-UI..bN W. L. Stewart-Senior Vice President - Power.
Attachment ec:
U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W.-
Suite 2900.
Atlanta, Georgia 30323 Mr. J. L. Caldwell
. NRC Senior Resident inspector North Anna Power Station i
1
ATTACHMENT EVALUATION OF NORTH ANNA POWER STATION UNIT 1 STEAM GENERATORS' CATEGORY. C C TUBE INSPECTION RESULTS l
VIRGINIA ELECTRIC AND POWER COMPANY l
L_--_________-____-__--_-___--_-_-_
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l l
Evaluation of i
North Anna Unit 1 Steam Generators'
]
Category C-3 Tube inspection Results 1
Introduction North Anna Unit 1 began commercial operatiors in June 1978 and has completed operation of its seventh fuel cycle. On February 25,1989, Unit 1 experienced a reactor trip from approximately 76% full power (the unit had been in a power coastdown). During the reactor trip event, a tube leak occurred in the 'C' Steam Generator. Based on the presumed scope of work to repair the steam generator tube leak, Virginia Electric and Power Company elected to commence the refueling outage originally scheduled to begin in April 1989. During the refueling outage, the cause of the steam generator leak was investigated and found to be a failed tube plug. Corrective actions related to the tube plug failure are the subject of another letter.
During this refueling outage, eddy current examinations of the steam generator tubes were conducted as required by the augmented eddy current examination program described in the facility Technical Specifications. The results of these steam generator tube eddy current inspections have been classified
)
as Category C-3 for each of the unit's three steam generators.
Inspections Performed During the 1989 refueling outage, an extensive eddy current inspection program was conducted in all three steam generators. This program relied on diverse, redundant inspection methods in areas of the steam generator tubing which had previously experienc1d tube degradation. The eddy current
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inspection methods utilized during the outage included bobbin coil testing,8x1 probe testing, rotating pancake coil testing, and profilometry.
For Steam Generators 'A' and 'B', the following inspections were conducted:
100% of the available tubes by bobbin coil probe
+
100% of the available tubes throgh the first support plate (hot leg) by 8x1 probe Tubes in rows 8 through 12 through the seventh support plate (hot leg and cold leg) by 8x1
+
probe Approximately 3% of the available tubes through the seventh support plate (hot leg) by profilometry examination l
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l L________________________-____-_____________--_-_--_____-_________-___-__-______________-_____-________-__--_-
c..
. For Steam Generator 'C', the following inspections were conducted:
- - 100% of the available tubes by bobbin coil probe 100% of the available tubes through the seventh support plate (hot leg) by 8x1 probe Tubes in rows 8 through 12 through the seventh support plate (cold leg) by 8x1 probe -
Approximately 3% of the available tubes through the seventh support plate (hot leg) by; profilometry examination For Steam Generators 'A', 'B', and 'C', the U-bend section of tubes in row 2 were inspected with a rotating pancake coil (RPC) probe (row 1 tubes had previously been plugged). In addition, distorted indications
. (Dis) identified by bobbin probe examination and'possible indications (Pis) identified by 8x1 probe
- examination were confirmed by RPC.
Technical Specification 4.4.5.2' requires that a sampling plan be developed as specified in Technical Specification Table 4.4-2. The required samples were selected prior to the inspection and results of inspections reviewed to determine the applicable Technical Specification Category (i.e., C-1. C-2, or C-3) after completion of the inspection.
The type of inspection performed, extent of inspection, reason for performing the inspection, and
' inspection findings for Unit 1 steam generators are summar.a J in Table 1 (attacheQ.
Insoection Results anci Corrective Actions Steam Generator 'C' was the first to be inspected. The inspection results were reviewed and determined to be in the C-3 category. Later in the outage, after the examination of the tubes in Steam Generc. tors 'A'
' q and 'B' had been completed, their inspection results were reviewed and also determined to be in the C-3 category. The NRC was promptly notified in accordance with 10 CFR 50.72 of the category C-3 determination for 'C' Steam Generator and notified again upon determination of the category C-3 for Steam Generators 'A' and 'B'. A description of the situation was reported to the NRC and is contained in a Licensee Event Report (LER) for Unit 1, LER 89-004-00, dated May 5,1989. Additional details of the investigation conducted are contained herein.
As a result of the eddy current inspections,39 tubes were required to be plugged in Steam Generator 'A'.
Of those,28 tubes exhibited cracking at a tube support plate (TSP) caused by primary water stress corrosion cracking (PWSCC) arrJ 1' iubes were plugged for indications of PWSCC at the expansion transition zone (EZ) near the top of the tubesheet (TTS).
_wx__--,.---_--_a---_---___-,----,.---__-------------,--_----_---.--------------_-__----_-------w---___-----,--------------,-----------,..----.----._u-----.----
For Steam Generator 'B',50 tubes were required to be plugged. The following list provides a distribution of plugging required:
+
8 tubes PWSCC at EZ 2 tubes Row 2 U-bend PWSCC.
1 tube TTS O.D.
- 2 tubes Anti-vibration bar (AVB) wear
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- O.D. origin in the sludge pile in addition to these pluggable tubes, we will install dampeners in 3 tubes as a conservative measure.
These dampeners are to be installed because of concerns related to the previous U-bend fatigue issue and will minimize the potential for developing fatigue cracking in the plugged tubes. Of these 3 tubes,2 tubes have never been plugged. Tubes with dampeners will be plugged with a sentinel plug in the cold leg and a solid mechanical plug in the hot leg. Therefore, the total new tube plugging for 'B' Steam Generator is 52 tubes for this outage.
Eddy current testing in Steam Generator 'C' resulted in the plugging of 56 tubes. The following list provides a distribution of plugging required:
42 tubes PWSCC at TSP 9 tubes PWSCC at EZ 2 tubes TTS O.D.*
2 tubes TSP O.D."
1 tube AVB wear O.D. origin in the sludge pile
" O.D. origin within tube support plates in addition to these pluggable tubes, we installed 7 cold leg sentinel plugs and 7 hot leg solid mechanical plugs in a " box" around tube location R3C60.1 This is the tube that contained the failed tube plug and in 1Eight tubes were required to be plugged to form a " box" around this tube location. One tube,R2C60, had been previously plugged with sta;1dard mechanical plugs on both the hot leg and cold leg sides. This tube will remain plugged with the standard mechanical plugs.
t which the plug top remains lodged in the U-bend portion of the tube. This corrective measure will prevent significant primary-to-secondary leakage should the plug top become dislodged and begin to wear on an adjacent tube. The plug top will be contained within the " box" and the seventh tube support plate.
Calculations establish that, should the plug top become dislodged from R3C60 and wear against one of the plugged tubes in the " box," through-wall penetration of that tube would not be expected in less than
- eight years of operation. The sentinel plugs installed on the cold leg will provide a low-leakage, early warning of this postulated event.
Finally in Steam Generator 'C', as an additional conservative corrective measure, we will install dampeners and plugs in 2 tubes. These 2 tubes have never bee:i plugged. The total new tube plugging for Steam Generator 'C' for this outage is 65 tubes.
The following table is a summary of tube plugging levels for Unit 1 steam generators:
A B
C Previously 314 217 380 Plugged 1989 Outage 39 52 65 Total Plugged 353 269 445
% Plugged 10.4 7.9 13.1 North Anna Unit 1 currently has a tube plugging limit of 18% for each steam generator.2 Based on previous plugging and known degradation mechanisms,1989 outage plugging totals are within the expected range. Adequate margin is available to permit future tube plugging and continued operation of these steam generators. Plugging tubes with confirmed defects returns the steam generators to an operable condition.
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2The 18% steam generator tube plugging limit is one of the major assumptions in our most recent LOCA-ECCS reanalysis transmitted to the NRC by letter Serial No.88-605, dated September 30,1988, and approved by the NRC by License Amendment No.114 for North Anna Unit 1, dated January 17,1989.
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Ooeratino Historv During the recently completed cycle of operation, the steam generators performed satisfactorily. Primary-to-secondary leakage was minimal. The observed leakage rate based on the condenser air ejector radiation monitor indications was 0.59 gallon per day (GPD). Based on N 16 monitors, the following leakage rates were observed at the end of the operating cycle:
Steam Generator 'A':
lessthan 1 GPD Steam Generator 'B':
2.4 GPD Steam Generator'C':
less than 1 GPD After returning to service in October 1987, no forced outages occurred due to primary-to-secondary leakage in excess of Technical Specification limits until the reactor trip and the subsequent tube leak event on February 25,1989. The secondary side chemistry remained within Virginia Electric and Power Company guidelines (which are consistent with industry standards) during the operating cycle.
Basis for Return to Service As previously discussed, extensive eddy current inspection using analysts trained to an approved rule base revealed indications of tube degradation. Tube plugging has removed the unacceptable tubes from l
l servica. The steam generators have been returned to an operable condition by plugging unacceptable tubes. Several additional tubes will be plugged as a conservative, preventiva measure. Plugs used for plugging the tubes discussed in this report were made from the preferred microstructure heat of material thereby assuring continued safe operation. The topic of plug failure is discussed in a separate report.
Eddy current inspection did not reveal a significant change in the rate of known degradation mechanisms.
No significant tube plugging from new degradation mechanisms has occurred. The existing degradation mechanisms are understood. The unit operated with little or no primary-to-secondary leakage since the October 1987 startup. In addition, based on previous observations, tube denting at tube support plates has been arrested. A statistical evaluation of tne profilometry data collected during this outage has not yet been completed, but no obvious increase in tube denting was observed. This !: attributable to the boric acid treatment program and the improved chemistry control during the operating cycle, in the longer term, we have begun initial efforts for the potential replacement of the Unit 1 Steam Generators. Planning has begun and the design specifications are prepared. Replacement is expected to occur no earlier than 1995. The replacement Steam Generators will contain Alloy 690 tubing and incorporate the latest l
features for reliable operation.
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