ML19309A741
| ML19309A741 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/28/1980 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton, Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8004010223 | |
| Download: ML19309A741 (8) | |
Text
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'fI Wisconsin Electnc eoun coursur 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE. WI 53201 March 28, 1980 Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation U.
S. NUCLEAR REGULATORY COMMISSION Washington, D. C.
20555 Attention:
Mr. A.
Schwencer, Chief Operating Reactors Branch #1 Gentlemen:
DOCKET NO. 50-266 RESULTS OF STEAM GENERATOR INSPECTION POINT BEACH NUCLEAR PLANT, UNIT 1 In accordance with the Nuclear Regulatory Commission's November 30, 1979, Confirmatory order for Modification of License in the above named docket, Point Beach Nuclear Plant Unit 1 was taken out of service on February 28, 1980, after 60 effective full power days of operation since its last refueling outage for hydrostatic tests and oddy current examination of the steam generator tubes.
The purpose of this letter is to present the results of this inspection, as required by the November 30 Order, and to request your written concurrence that the results of these tests are acceptable so that Unit 1 may be returned to service.
Although Point Beach Unit 1 was taken out of service on February 28, 1980, the primary-to-secondary and secondary-to-primary hydrostatic leak check of both steam generators were not conducted until March 6 because of concurrent steam generator i
repair and inspection work underway at Point Beach Nuclear Plant Unit 2.
The results of the Unit 1 hydrostatic leak checks revealed no leakage in the "A" steam generator and one tube (R23C44) which exhibited a slight leak at a rate of three drips per minute, and one wet plug in a previously plugged tube (R23C50) in the "B"
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Following completion of eddy current inspection of Point Beach Unit 2, the eddy current equipment was transferred l
to Unit 1 on March 10 and eddy current examination of the s
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. Unit 1 steam generators commenced on March 11.
The eddy current inspection was conducted in accordance with the eddy current 4%
I program submitted with our letter dated February 26, 1980, k O'
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4-Mr. liarold R.
Denton March 28, 1980 except that, as requested by your Staff, a minimum of 3% of the cold leg tubes were inspected in each steam generator from the hot leg side.
The inspection of the Unit 1 steam generators was completed on March 14.
Approximately 1,000 tubes were inspected in each steam generator.
These totals included over 100 randomly selected tuben in each steam generator which were inspected over their entire length.
The results of the eddy current inspection are presented
'in Attachment 1.
These resulta may be nummarized as follows:
"A" SG "D" SG Deep Crevice Defecta 18 23 Undefinable Indicationo 6
3 (located within tubesheet)
Defects Slightly Above 0
1 the Tubcoheet t
Leaking or Dripping Tubco 0
1 (no defect found) 24 28 These results were reviewed with members of your Staf f at Point Beach on March 13 and 14, 1980.
Following these discussions, we agreed to pull three tubeu from the "B" steam generator to perform additional metallurgical examinations of f
the defects identified in thene tuben.
The tuben which were pulled are identified as R19C37, R30C41 and R26C53.
The tube pulling efforta commenced on March 17 following.explonive tube plugging of the 24 tubes in the "A" nteam generator and 25 tuben in the "B" steam generator.
Several difficultion were encountered during the tube pulling and the last of the three tubes was not removed until March 23.
Weld repairs were completed on the pulled tube holen on March 28.
Because of the problema encountered while pulling tube R30C41, an additional four tubes adjacent-to thin tube were plugged as a precautionary measure resulting in all tubes surrounding R30C41 being plugged.
Thus, a total of 32 tubes were removed from nervice in the "B"
As of March 26, the tube pulling and tube weld repair activities had resulted in approximately 145 man-rom of e;;pocure.
In addition to thcae activities, the eddy current and vinual inspections, the tube plugging and the general health phynica coverage have contributed another 44 man-rem exposure.
It in expected that by the time the weld repairs have been completed on March 28, approximately 10 man-rom additional exposure will be recorded, i
Mr. Harold R.
Denton March 28, 1980 We have completed preliminary inspections and metal-
'ography on samples of the removed tubes.
The preliminary results of the evaluation of these samples were provided to members of your Staff during a meeting at the Westinghouse Forest Hills Laboratory on March 28, 1980.
The November 30 Confirmatory Order also required that we provide two additional items together with the summary of the results of the eddy current examination.
These items are the photographs of the tubesheet of each steam generator af ter plugging and a description of the quality assurance program in use for verification of tube examinations and tube pluggi;.g.
This information is provided in Attachments 2 and 3, respectively, to this letter.
In addition to the information provided to your Staff on March 28, we will also provide the following by April 30, 1980:
1.
A detailed report on the metallographic, chemical and laboratory examinations performed on the removed tubes; 2.
A detailed re-evaluation of all previous eddy current tapes for tubes R19C37, R22C46, R30C57, R28C38, R30C44, R32C42, R26C53, and R30C41; 3.
Information on the capability of re-evaluating 400 KHZ single frequency eddy current signals by mixing with a psuedo tubesheet or tube support plate signal; and 4
4.
A section or sections of the removed tubes to a NRC consultant for chemical and metal-lurgical examination, if requested by the NRC.
During the 60 effective full power days of operation preceding the current shutdown, Point Beach Unit 1 experienced a slight indication of primary-to-secondary leakage of about 30 l
gallons per. day.
This leakage rate, which remained constant over the 60 EFPD period, may be indicative of a slowing of the rate of corrosion in the tubesheet crevice.
Similarly, the hydrostatic leak checks in each steam generator at the end of thin 60 EPPD period indicated no increase in' tube leakage.
The preliminary results of the laboratory examination of the tube specimens which have been removed from Unit 1 during this outage show the same general condition as tubes removed during October 1979 with regard to general intergranular attack in the tubesheet crevice.
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Mr. Harold R.
Denton March 28, 1980 The examination methods included eddy current' examination, X-radiography and examination of :aetallographic sections showing material microstructures at various locations in each tube.
-However, since eddy current exami. nations have detected some indications in the crevice region, a program to provide assur-ance of safe operation will be continued.
We believe that the following conditions are appropriate to demonstrate the continued safe operation of Point Beach Unit 1:
1.
Within 90 EFPD, a 2,000 psid primary-to-secondary hydrostatic test and a 800 psid secondary-to-primary hydrostatic test will be performed.
An eddy current examination consisting of about 1,000 tubes in the central region of the hot leg in each steam generator and 3% of the remaining tubes outside this area will be performed.
2.
Primary coolant activity for Point Beach Unit 1 will be limited in accordance with the provisions of Sections 3.4.8 and 4.4.8 of the Standard Technical Specifications for Westinghouse Pressurized Water Reactors, Revision 2, July 1979, rather than Technical Specification 15.3.1.C.
3.
Close surveillance of primary-to-secondary leakage will be continued and the reactor will be shutdown for tube plugging on confirmation of any of the following conditions:
'I a.
Primary-to-secondary leakage of 150 gpd (0.1 gpm) in either steam generator; b.
Any primary-to-secondary leakage in excess of 250 gpd (0.17 gpm) in either steam generator; or c.
An upward trend (average over a three-day period) in primary-to-secondary leakage in either steam generator in excess of 15 gpd (0.01 gpm) per day, when measured primary-to-secondary leakage is above 150 gpd in that steam generator.
4.
The reactor will be shutdown, any leaking steam generator tubes plugged, and an eddy current examination as described in Item 1.,
t-above, will be performed if leakage due to
Mr. Harold R.
Denton March 28, 1980 crevice corrosion in either steam generator exceeds the limits stated in Technical Specification 15.3.1.D.
5.
Unit 1 will be operated at a reactor coolant pressure of 2,000 psia with the associated parameters (i.e., overtemperature AT and low pressurizer pressure trip point) with the limits indicated in the Safety Evaluation Report appended to your letter of January 3, 1980.
On return to power operation, we will continue the following program which we believe will assist in retarding further tube degradation:
a.
Unit I will be operated at a reduced reactor coolant system hot leg temperature.
b.
Continue close surveillance of feedwater chemistry conditions and condenser tube leakage.
c.
Perform sludge lancing within nine months of returning to power.
We believe that Unit 1 will be ready to return to power by March 31, 1980, and would appreciate your continuing immediate review and reply regarding this matter.
Please telephone us if you have any questions or require further information.
Very truly yours,
?
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Sol Burstein Execu ive Vice President
/
Enclosures Copies to Mr. L. T. Mittness (PSCW)
Mr. Peter Anderson (WED)
Ms. Joan Estes (LCFSE) 1 I
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d' TABLE 1 "A"
STEAM GENERATOR TUBES PLUGGED 7
Tube Defect Location J
R12C19 80%
19-21" above tube end P07C22-29/96%
12/17" above tube end R18C22 66%
20" above tube end R10C23 41t 20" above tube end R07C24 83%
17-20" above tube end 17-21" above tubo end R08C24 79%
R25C45 69%
12-20" above tube end R20C48 85%
21" above tube end R09C49 90%
21" above-tube end R17C50 85%
19" above tube end R19C50 97%
11" above tube end R20C50 97%
11" above tube end R12C59 87%
21" above tube end R12C61 83%
17" above tube end R14C63 83%
19" above tune end R15C66 60%
18" above tube end R20C41 91%
19" above tabe end 4
R25C43 73%
17" above tube end R15C28.
Undefinable Indication 21" above troe end R28C34 Undefinable Indication 18-20" above tube end R28C35 Undefinable Indication 17" above tube end i
- RllC46 Undefinable Indication 12-21" above tube end R29C52 Undefinable Indication 14" above tube end R08C27 Undefinable Indication 15-20" above tube end f
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ATTACHMENT 2 censists of eight photographs covering the Point Beach Nuclear Plant Unit 1 "A" and "B"
steam generator inlet and outlet tubesheets.
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RTTACHMENT.3
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4 DESCRIPTION OF QUALITY ASSURANCE PROGRAM FOR TUBE EXAMINATION AND PLUGGING Procedures are in use at the Point Beach Nuclear Plant to verify the identity of steam generator tubes selected for repair or removal.
These procedures require the reexamination lif all~ tubes designated as requiring repair following evaluation of data collected during the normal eddy current examination.
The specific tube requiring repair is hand-probed and the eddy current signals monitored for comparison with the photographic record of the individual indication from the initial tube examination and evaluation.
The information from this hand-probing is also recorded on magnetic tape and a strip chart recorder.
When the eddy current operator is satisfied that the observed signal matches the-photographic record, the operator located outside the steam generator channel head is instructed to physically mark the tube, using chalk penetrant developer, plastic plugs or an equivalent method to positively identify the tube to be plugged.
After the contractor has identified and marked all tubes to-be plugged, a member of the Plant staff independently checks the tubesheet against the list of tubes to be plugged and signs off that the correct tubes have been marked.
Following this sign-off, the tubes are ready to be plugged.
The operator outside the channel head describes and points out the tubes to be plugged to the jumper, the person who will actually enter the channel head and place the plugs.
The jumper enters the steam generator, places the plugs, and leaves the steam generator as quickly as possible in order to minimize personnel exposure.
Before the tube plugs are detonated, the operator looks into the steam generator to make sure the plugs are correctly placed and have not slipped.
After all tubes have been -plugged, a Plant staf f member again checks the tubesheet against the list of tubes which had to be plugged and signs off that the proper tubes were in fact plugged.
A photograph of the tubesheet is taken to preserve a record of all tubes currently plugged.and unplugged.
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