ML20041B505
| ML20041B505 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/15/1982 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton, Novak T Office of Nuclear Reactor Regulation |
| References | |
| TAC-48122, NUDOCS 8202240148 | |
| Download: ML20041B505 (14) | |
Text
1 i
Wisconsin Electnc eona couraur 231 W. MICHIGAN, P.O. BOX 2046 MILWAUKEE. WI 53201 February 15, 198
/g C
Mr.
H.
R. Denton, Director rgrgG;
~3 FEB 2 S 1992> r$
Office of Nuclear Reactor Regulation p
p U. S. NUCLEAR REGULATORY COMMISSION grgy *, 7 );
Washington, D. C.
20555 2..
Attention:
Mr. Thomas M. Novak Q
Division of Licensing
'?
Gentlemen:
DOCKET NO. 50-266 STEAM GENERATOR INSPECTION AND OPERATING INTERVAL POINT BEACH NUCLEAR PLANT, UNIT 1 On January 27, 1982, we met with the NRC Staff to review our most recent experience with Point Beach Nuclear Plant, Unit 1, steam generators and to provide our response to the Staff's evaluations of this experience which were included in Mr. Novak's letter of December 30, 1981.
At the conclusion of this meeting, Mr. Novak requested that we summarize in writing the information presented and discussed during the meeting and that we provide additional information regarding defects which were detected at or above the top of the tubesheet.
This transmittal is in response to those requestc.
Attached hereto are enclosures covering the various items of our discussions as follows:
1.
Summary of Crevice Corrosion Indications provides a summary of eddy current indications in steam generators A and B since August 1979.
The table in Enclosure 1 has been corrected for av error in addition in the " cumulative total" column for steam c
generator B.
This correction does not affect the evaluations presented during the meeting.
/bo/
e
If f I t 8202240148 820215 PDR ADOCK 05000266 G
Mr. H.
R. Denton February 15, 1982 As discussed during the meeting and reported in Licensee Event Reports, periodic steam generator inspections have continued to detect small volume indications of crevice corrosion in both steam generators which were not identified during previous inspections.
Comparison of eddy current tapes from a given inspection to tapes from prior inspections allows an estimate of the rate of progression of crevice corrosion after assignment of indications to the time of first appearance on eddy current tapes.
The summary provided in Enclosure 1 demonstrates that crevice corrosion continues to progress in both steam generators but that the rate of appearance of new defects has been significantly reduced since lowering reactor coolant temperature following the October 1979 refueling outage.
The rate of progression also appears not to have changed in either steam generator since raising reactor coolant temperature from 557'F to 575'F in July 1981.
2.
Rate of Progression of Crevice Corrosion provides a graphical summary of data in.
Since the rate of crevice corrosion is believed to be dependent on temperature, the cumulative total of crevice indications is plotted versus the approximate time at hot conditions rather than calendar time. has also been corrected for the minor addition error in Enclosure 1.
As we discussed, Enclosure 2 demonstrates that reduced reactor coolant temperatures have significantly reduced the rate of progression of crevice corrosion in both steam generators.
The rate since about July 1980 l
appears to be relatively constant and does not appear to have increased since raising temperatures in July 1981.
As we also discussed, we would not expect the rate of a temperature-dependent corrosion process to decrease continually unless the environment is modified by removal of corrosive contar. nants.
Since 1979, we have performed crevice cleaning procedures during scheduled shut-downs and have removed significant quantities of c
residual sodium and phosphate from both steam generators.
The effects of this cleaning procedure may be reflected in the generally decreasing rate of progression prior to about July 1980 and the continuing low rate of progression since July 1980.
Mr. H. R. Denton February 15, 1982 3.
Crevice Cleaning Experience is a summary of estimates of the average quantities of sodium and phosphate removed from Unit 1 steam generators per cleaning cycle and indicates a generally decreasing trend in the quantities removed.
This experience is as expected since sodium phosphates are not added to the steam generators but are present as residuals from earlier use of phosphate chemistry.
As we discussed, the use of the crevice cleaning proce-dure may serve not only to remove contaminants from the steam generator crevices but also to modify the relative concentrations of the contaminants within the crevices.
Thus, even though contaminants may not be completely removed, the formation of free caustic, which is believed to be the principal cause of crevice corrosion, may be minimized.
4.
Comparison of Eddy Current Signals provides a summary of comparisons of the October 1981 eddy current indications with July 1981 and December 1980 eddy current tapes which were provided in LER 81-017/0lT-0.
While 17 tubes were found to contain degradation in excess of the plugging limit in October 1981, only seven of these were new indications which were not present in July 1981.
The comparisons provided in pages 3 and 4 of the attachment to LER 81-017/0lT-0 show that, for steam generator A, 31 eddy current indica-tions were identified on the hot leg.
Of these, six were identified as being at or above the top of the tubesheet.
Of the six indications, two exhibited some change and four exhibited no change since July 1981.
Comparing the July 1981 indications to the December 1980 indications for the same tubes shows that one exhibited some change, four exhibited no change, and one was not present in either inspection.
Similarly, for the 25 crevice indications in steam generator A, ten exhibited some change since July, 14 exhibited no change, and one was not tested in July.
Comparing July 1981 indications to December 1980 indications for these 25 tubes shows that 15 exhibited some change, six exhibited no change, three were not present in either inspection, and one q
was not tested in July.
Similar comparisons for steam generator B were also discussed.
All comparisons are summarized in Enclosure 4.
The number of tubes exhibiting signal changes since July 1981 is less than the number
Mr. H. R. Denton February 15, 1982 which exhibited signal changes between December 1980 and July 1981.
This experience does not suggest an increase or decrease in the rate of corrosion in either steam generator as a result of increasing reactor coolant temperature in July 1981.
5.
Steam Generator Leakage Experience provides a summary of Unit 1 steam generator leakage experience and leaking tubes or plugs since December 1979.
As we discussed, this experience has shown consistently low levels of primary-to-secondary leakage due primarily to leaking explosive plugs.
In particular, there have been no forced outages of Unit 1 since December 1979 due to steam generator leakage.
In addition, there have been no leaking tubes in either steam generator since July 1980.
This experience suggests that even though crevice corrosion is continuing, it is continuing at a very low rate and is not likely to result in significant numbers of leaking tubes even with increased reactor coolant temperature.
We also discussed our. leakage experience on Unit 1 since returning to power after the fall 1981 refueling outage.
As indicated in Enclosure 5 and reviewed during the meeting, the initial primary-to-secondary leakage was in the order of ten gallons per day.
This leakage was believed to be due to leaking or " wet end" explosive plugs which were not repaired during the refueling outage due to personnel exposure considerations.
The leakage rate increased gradually thereafter to approximately 20 gallons per day and has remained essentially constant at that level since mid-January 1982.
These levels of primary-to-secondary leakage are not significantly different from levels experienced in the past.
- However, as we indicated during our discussions, we shall continue to monitor the leakage rate and will consider a shut-down for repair if the leakage rate indicates a continued upward trend which could be expected to exceed 50 to 100 gallons per day.
6.
Safety Significance We also discussed the safety significance of continuing crevice corrosion and our recent experience and agreed that our proposed full cycle operating and inspection interval has no adverse safety implicction, given the previous extensive evaluations performed by the Staff and the conservative leakage rate limits in the Unit 1 i
i
Mr. H. R. Denton February 15, 1982 Technical Specifications.
The " Safety Evaluation Report Related to Point Beach Unit 1 Steam Generator Tube Degradation Due to Deep Crevice Corrosion, November 30, 1979" (SER) considered, among other, matters, the safety significance of crevice defects and concluded at page 24:
"3.
Conservative primary to secondary leak rate limits will provide assurance that in the event that large defects go undetected, or the corrosion rate accele-rates, timely plant shutdown and corrective actions can be taken."
The SER also concluded at page 25:
" Finally, even if a few tubes went undetected by ECT and hydrotests, became severely degraded without leaking, and collapsed during a postulated LOCA, the resulting in-leakage would be tolerable because of the collapse failure mode and the large leak rates required to adversely affect ECCS performance."
We agreed that these conclusions by the Staff remain valid.
We also discussed the estimated leakage and number of tube failures required to adversely affect ECCS performance and concluded, on the basis of estimates in Appendix A of the SER and data provided with our letter of March 4, 1981, that more than 100 tubes would be required to fail simultaneously in the crevice during a loss-of-coolant accident to even begin to affect ECCS performance.
.Our experience since December 1979, as indicated in Enclosures 1, 4, and 5, and hydrostatic tests have shown this will not occur.
We also reviewed briefly the' experience with multi-frequency eddy current testing and the occasional detection of degradation at or above the top of the tubesheet.
As in the past, these indications are attributed to thinning or cracking in prior years and it is not unexpected that we continue to detect small volume indications with the use of refined eddy current techniques.
These indications have no significant safety implications and there is no evidence of significant corrosion occurring at or above the top of the tubesheet. is an updated tube plugging history for Unit 1 and demonstrates the lack of significant corrosion in regions outside the tubesheet crevice.
Mr. H. R. Denton February 15, 1982 In response to Mr. Novak's request, Enclosure 7 provides a summary of eddy current signal comparisons reported in LER's since October 1979 for tubes with indications at or above the top of the tubesheet.
In
~
some cases, the eddy current evaluation for a particular tube resulted in different interpretations during subsequent comparisons.
Where this occurred, both values are reported.
Inspection of Enclosure 7 shows a lack of significant progression of corrosion at or above the top of the tubesheet.
This has been verified in the past by destructive examination of tubes removed from Unit 1 steam generators and by comparison of eddy current signals during outages prior to October 1979.
In summary, we believe that our proposed operating and inspection intervals are appropriate for the following reasons:
1.
The apparent progression of crevice corrosion remains at a reduced rate even after raising reactor coolant temperature in July 1981.
2.
Crevice cleaning procedures and operation at reduced temperature appear to have been at least partially effective in modifying the crevice environment such that corrosion can be maintained at an acceptably low rate for the full operating cycle.
3.
There have been no forced outages of Unit 1 due to tube leakage since December 1979 and no leaking tubes in either steam generator since July 1980.
Primary-to-secondary leakage has been very low since 1979.
4.
There is no evidence of significant corrosion at or above the top of the tubesheet.
5.
There are no safety concerns regarding the proposed operating and inspection intervals.
The Staff evaluations in the SER remain valid regardless of the length of these intervals.
6.
Even if some tubes were to experience corrosion to the point of leaking, the conservative leak rate limits in the Technical Specifications ensure timely plant shut-down and corrective action.
Mr. H.
R. Denton February 15, 1982 We trust that this information which was discussed at the January 27 meeting is responsive to the concerns expressed in your December 30 letter.
We will, of course, continue to monitor the primary-to-secondary leakage in Unit 1 and will consider a shut-down for inspection in the event that a continued increase is experienced.
We believe that this course of action is apprcpriate and in keeping with our normal safe operating practice, We shall, of course, keep you fully informed of any modifications to our operating plans.
Very truly yours,
-A Exe utive Vice President Sol Burstein Enclosures Copies to NRC Resident Inspector C. F. Riederer (PSCW)
Peter Anderson (WED) i l
i 1
i
ENCLOSURE 1
~
POINT DEACil NUCLEAR PLANT, UNIT 1
SUMMARY
OF CREVICE INDICATIONS
- AFTER ADJUSTMENT FOR TIME OF FIRST DETECTION STEAM GENERATOR A Unidenti-Cumulative Outage fiable 100%
90-99%
80-89%
70-79%
60-69%
50-59%
<50 Total Total Prior to and including August 1979 4
8 20 28 6
2 1
69 69 October 1979 2
1 38 30 13 4
2 10 100 169 December 1979 7
1 7
2 2
8 27 196 March 1980 3
4 6
4 4
4 6
31 227 July 1980 3
8 5
3 9
5 33 260 1
2 16 276 December 1980 13 July 1981 4
1 1
9 285 1
2 October 1981 1
2 4
289 1
STEAM GENERATOR D Prior to and including August 1979 1
17 22 7
4 51 51 October 1979 4
2 27 19 16 6
3 14 91 142 19 161 December 1979 2
1
-4 8
3 1
March 1981 2
1 4
3 1
5 4
20 181 July 1980 2
1 2
10 8
10 2
1 36 217 Dccember 1980 1
1 2
219 July 1981 1
1 220 October 1981 1
1 1
3 223
- All indications are included.
Since some tubes had more than one indication and/or indications less m.
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4 ENCLOSURE 4 1
COMPARISON OF EDDY CURRENT SIGNAL CHANGES FOR TUBES WITH HOT LEG INDICATIONS IN OCTOBER 1981ti)'
October 1981 Compared To July 1981(2)
At Or Above Tubesheet Crevice Change No Change Change No Change Steam Generator A 2
4 10 14 Steam Generator B 0
3 3
5 TOTAL 2
7 13 19 July 1981 Compared to December 1980(2, 3)
Steam Generator A 1
4 15 6
Steam Generator B 1
2 3
2 TOTAL 2
6 18 8
x (1) Based upon data provided in LER 81-017/01T-0 (2) One tube not tested in July 1981, steam generator A.
(3) Four tubes in steam generator A and three tubes in steam generator B showed no defects in either inspection.
t
t STEAM GENERATOR LEAKAGE IIISTORY POINT BEACil NUCLEAR PLANT, UNIT 1 Prior STEAM GENERATOR A STEAM GENERATOR B Date of Leak Leaking Leaking Wet Leaking Leaking Wet Shutdown Rate Tubes Plugs Plugs Tubes Plugs Plugs (gpd) 12/79 (1) 0 0
0 1
2 0
3/80 20-34 0
0 0
1 0
1 7/80 10-20 0
2 2
1 0
1 11/80 4-10 0
0 2
0 0
2 7/81 4-6 0
3 3
0 1
2 10/81 5-10 0
0 0
0 1
2 Present 10-20 M
5 vs (1) Forced outage due to tube leak.
M un i
UNIT 1 STEAM CENERATOR TOBE PLUGGIhG HISTORY Tube 1s Pluggrd Data of Thinning or Crevice Cumulative Ou tage Elapsed Time Denting Cracking Corrosion Other Total Percent (Years)
A, 8
A B,
A, B
A B
A, B
A, B
12/21/70 0
1(1) 1
<0.1 0
9/30/72 1.8 87 91 14 4(2) 102 95 3'.1 2.9 j
4/6/74 3.3 1
1 103 96 3.2 2.9 2/26/75 4.2 59 98 162 194 5.0 6.0 11/16/75 4.9 6
4 168 198 5.2 6.1 10/1/76 5.8 168 198 5.2 6.1 6/24/77 6.5 1
168 199 5.2 6.1 10/4/77 6.9 10 1
2 179 201 5.5 6.2 2/1/78 7.1 1(3) 180 201 5.5 6.2 5/26/78 7.4 1
181 201 5.5 6.2 9/20/78 7.7 1
6 4
188 205 5.7 6.' 3 3/1/79 8.2 8
1 196 206 6.0 6.3 8/5/79 8.6 52 45 248 251 7.6 7.7 8/29/79 8.7 2
2(4) 252 251 7.7 7.7 10/5/79 8.8 2
3(6) 68 61 7
4(5) 329 319 10.1 9.8 12/11/79 9.0 19 15 1(7) 349 334 10.7 10.2 2/28/80 9.2 1(8) 24 26 9(9) 373 370 11.4 11.3 28 22 3(10) 404 392 12.4 12.0 7/26/80 9.6 11/26/80 9.9 3
7 407 399 12.5 12.2 7/4/81 10.5 1(8) 2 2
409 402 12.5 12.3 10/9/81 10.8 9
7 412(11) 409 12.6 12.5 Notes:
(1) Plugged during manufacture.
(2) Fourteen tubes in A were plugged due to gouging during machining for clad repair. Three tubes in B were removed for analysis and one was plugged by mistake.
(3) Plugged tube was in periphery.
(4) An audit of tubesheet photographs indicated two tubes which were plugged but previously not included in inspection reports.
(5) Seven tubes in A included three with defects less than the plugging limit, two tubes which had no indications but which were pulled for analysis, and two tubes plugged by mistake. Four tubes in B included three tubes with indications less than the plugging limit and one tube plugged by mistake.
n (6) Two tubes in A and three tubes in B were plugged due to defects identified at or above the tubesheet using t*
multi-frequency eddy current techniques. These defects are attributed to thinning or cracking in prior O
years, based upon comparison with single-frequency eddy current results from previous inspections.
y (7) One tube plugged by mistake.
m (8) One tube in B was plugged due to a defect above the tubesheet which was identified using multi-frequency to techniques. This defect is attributed to thinning or cracking in prior years, based upon comparison with results from previous inspections.
(9) Four tubes in B were plugged due to the possibility of being damaged during tube pulling operations and five leaking tubes were plugged without identifying the defect location.
(10) Three tubes plugged by mistake.
(11) One tube which was in excess of the plugging limit was repaired by sleeving. Plugs were removed from six tubes and the tubes were sleeved and returned to service.
ENCLOSURE 7 POINT BEACil NUCLEAR PLANT, UNIT 1
SUMMARY
OF EDDY CURRENT INDICATIONS STEAM GENERATOR IIOT LEG AT OR ABOVE TOP OF TUBESilEET t
TUBE NO.
10/81 07/81 12/80 07/80 03/80 12/79 10/79 STEAM GENERATOR A R05C69
<20%
NDD NDD R05C68 21%
NC NC R06C81
<20%
NC NC R10C54
<20%
ND, 32%
<20%
34%
34%
34%
NDD, 34%
R33C54 38%
DS, 20%
22%,
20%
34%
NT NT 34%
R36C29 38%
NC NC 58% plugged R19C37 STEAM GENERATOR B R27C30 25%
28%
29%
29%
NT NT R14C40 29%
28%
32%
R26C42 29%
21%, 29%
NDD, 29%
NDD - No detectable defect NC - No change when compared to 10/81 indication NT - Not tested this outage DS - Distorted signal
- - No comparison made for these inspections I
.