ML20028B415

LER 82-017/01T-0:on 821030,four Tubes in Steam Generator a & Three Tubes in Steam Generator B Indicated Degradation Greater than 40% Plugging Limit.Caused by Intergranular Attack.Tubes Mechanically Plugged

ML20028B415
Person / Time
Site:	Point Beach
Issue date:	11/15/1982
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML20028B410	List: ML20028B409 ML20028B415
References
LER-82-017-01T, LER-82-17-1T, NUDOCS 8211300363
Download: ML20028B415 (5)

Text

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NRC F ORM 366 U. S. NUCLE A',3 REHULATrJ.Y COMMISSION 87-7 1)

, LICENSEE EVENT REPORT

. CONTROL BLOCK: l l l l l l l (PLEASE PRINT G1 TYPE ALL REOUl%ED INFORMATION) 1 6 7

0 t 8

l9 Wl LICENSEE I l P l CODE B l H l141 l@150 l 0l -l LICENSE 0 l 0 lNUUSER 0 l 0 l 0 l -l 0 25l 0 lg 26 4 lLICENSE 1l 1 lTYPE 1l 1l@l JO l

57 CAT 58 lg CONT o 1 5$ng l L l@l 0l 5 l 0 l 0 l 0 l 2l 6 l 6 l@l 1 l 0 l 3 l 0 l 8 l 2 Ql 1 l 1l 1 l 5l 8 l 2 l@

7 8 60 61 DOCK ET NUMBER 68 69 EVENT DATE 74 75 REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h o 2 lunit 1 was shut down on 10/22/82 for refueling. Eddy current examina- l 0 3 l tion of the steam generator tubes was conducted from 10/26/82 to 10/30/83

, , ,On 10/30/82 verification of all initial eddy current data for tubes Four tubes

, , ,with indications exceeding the plugging limit was completed.

O 6 l l

,had indications greater than 40%. This event is similar to others and is reportable per Technical Specification 15.6.9.2.A.3.

7 3 g 80 DE CO E S 8C E COMPONENT CODE SUB DE S E O 9 l C l B l@ (_E_j@ W@ l H l T l E l X l C l H l@ W@ W @

7 8 9 10 11 12 13 18 19 20

_ SEQUENTIAL OCCUR RENCE REPORT REVISION LER'RO EVENT YEAR REPORT NO. CODE TYPE y g

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% 29 22 23 24 26 27 28 29 30 31 32 AKEN ACTO ON PL T MT HOURS S8 IT FOR b 8. SUPPLIE MANUFACTURER l_gj@34[_gj@

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42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h i O l All indications areater than 40% found during this inspection were withi-i i l in the tube sheet region and are considered to be IGA caused by causticj.

g All defective tubes identified during this inspection were mechanically, gl plugged. Operation of the unit at a reduced temperature will continue ,

, 4 g to minimize further IGA.  ;

7 8 9 80 ST S  % POWER OTHER STATUS IS RY OVERY DESCR QN i s W@ l 0 l 0 l 0 lgl N/A l [CJgl Eddy curreEt examinEtio l

/CTIVITY CO TENT RELEASED OF RELE ASE AMOUNT OF ACTIVITY LOCATION OF RELEASE i 6 [_z_j @ QlN/A l lN/A l 7 8 9 to 11 44 45 80 PERSONNEL EXPOSURES NUMBE R TYPE DESCRIPTION i 7 l 010 l 0 l@@@l N/A i l

PERSONNE L INJURIES

' NUMBER DESCRIPTION 7

i R 8 9 l 01 0 l 0l@l 11 12 N/A 80 l

LOSS OF OR DAMAGE TO FACILITY TYPE DESCRIPTION

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l 80 PUBLICITY ISSUE DESCRIPTION

~~s2113OO363 80 i15 NRC USE ONLY PDR ADOCK 05WO266 7

2 o 8 9 W [N/A to g PDR l 68 69 lllllllllllll7 80 3 C* W. Fay PHONE:

414/277-2811 o NAME OF PREPARER r, .- .- . ~_ ,, __

ATTACHMENT TO LICENSEE EVENT REPORT NO. 82-017/0lT-0 Wisconsin Electric Power Company Point Beach Nuclear Plant, Unit 1 Docket No. 50-266 On October 22, 1982, Unit 1 was shut down for its tenth refueling outage. The 2000 psid primary-to-secondary hydrostatic test condition was established during cooldown of the unit. An 800 psid secondary-to-primary leakage check was performed on both

} steam generators on October 26, 1982. The 800 psid secondary-to-primary leakage check was performed visually with the aid of remote video equipment. The specific conditions identified during the leakage checks are noted below. (All noted leaks were observed from previously plugged tubes.)

A" Steam Generator Hot Leg R 3C 9 Boric acid coated R12C25 Boric acid coated R14C57 2.0 drops per minute R31C31 1.5 drops per minute B" Steam Generator Hot Leg R13C63 2.0 drops per minute R28C43 Wet end R29C34 Boric acid coated R29C37 Boric acid coated R31C44 Wet end i

The eddy current inspection progr en, performed this outage, consisted of the following:

1. Inspection of essentially all readily remotely access-ible tubes to the first support plate in the hot legs of both steam generators.

2. Inspection over the U-bend from the hot leg side of the greater than 3% of the tubes in each steam generator.

3. Inspection up to the sixth support of the hot leg tubes containing sleeves.

4. A special inspection of all the sleet,es in both the hot leg and cold leg.

5. Inspection of tube locations previously identified as containing degradation.

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l On October 30, 1982, verification of all initial steam generator eddy current data for tubes with indications exceeding the plugging limit was completed. Four tubes in the "A" steam generator and three tubes in the "B" steam generator were verified to have degradation greater than 40%, which is the plugging limit of Technical Specification 15.6.2.A.5.

Of the 2,809 open tubes in the "A" steam generator, 2,769 were inspected and 2,787 of the 2,837 open tubes in the "B" steam generator were inspected. The tubes that were not inspected are as follows:

Number of Tubes Not Inspected Reason for Not Inspecting "A" "B" 19 16 Contained template plugs 1

1 -

Restricted tube ends 20 34 Under fixture " foot" 40 50 These tubes were not inspected because of the radiation i exposure associated with moving template plugs, moving eddy current i

equipment, and the reworking of a restricted tube end. The i

noninspected tubes constitute less than 2% of the unplugged tubes, most are not located in the zones where large numbers of defects have occurred, and the overall eddy current results did not indicate the necessity to inspect the tubes. Following is a summary of the l eddy current indications and comparisons with the data from the

! three previous eddy current inspections. A blank entry under the results of previous eddy current inspections in the following table indicates.that.the tape for that specific inspection was not examined for this comparison.

"A" Steam Generator l Hot Leg Tube Defect Location 03/82 10/81 07/81~

, R13C48 89% 21" ATE 89%.NC NDD

! R21C48 91% 20" ATE 90% NC NDD R19C56 83% 21" ATE 73%-C-DS 75-C-DS NDD l R27C58 80% 9&ll" ATE NDD R36C29 34% TTS R SC68 <20% 1/2" ATE l

R SC69 <20% 1/2" ATE R 6C81 <20% 1" ATE I

i

Cold Leg Tube Defect Location 03/82 10/81 07/81 4

R26C53 34% 2" ATS R20C60 <20% 5" ATS R28C48 36% 2" ATS "B" Steam Generator Hot Leg Tube Defect Location 03/82 10/81 07/81 R27C30 <20% 1" ATS R21C59 56% 8" ATE NDD R20C61 80% 20" ATE NDD R18C68 69% 20" ATE 73%-NC UDI UDI ATE - Above tube end NDD - No defect detected UDI - Undefinable indication ATS - Above tubesheet TTS - Top of tubesheet NC - No change C - Change DS - Distorted signal Four tubes in the "A" steam generator and three in the "B" steam generator contained indications exceeding the 40% plugging limit. Of the seven indications exceeding the plugging limit, one is a new indication in the "A" steam generator and two are new indications in the "B" steam generator. The other indications identified were either previously noted as undefinable indications or defects that previously existed, but were not identified in prior inspections. As in the past, all indications were small volume and originated on the tube's outside diameter.

The seven tubes containing indications greater than the plugging limit have been mechanically plugged. Correct plugging was independently verified by visual means.

In addition, the eddy current inspection program identi-fled a total, for both steam generators, of 71 tubes that are restricted to a 0.720" diameter probe at the first support plate on the hot leg side. Of the 71 restrictions, 20 are in the "A" steam generator and 51 are in the "B" steam generator. The majority of these restrictions are located in the periphery tubes near the

" wedge" areas.

J

e All of the restrictions, except for the two in the "B" steam generator, passed a 0.650" diameter probe. The two restric-tions noted above passed a 0.610" diameter probe. Thirteen of the twenty restrictions in the "A" steam generator were present during previous inspections. Forty-two of the fifty-one restric-tions in the "B" steam generator were present during previous inspections. In addition to a slight increase in the total number of restrictions, a slight increase in the extent of the restrictions in some of the tubes previously noted as containing restrictions was experienced.

All tubes restricted to a 0.700 probe at the first support were probed through the sixth support with a 0.650" or 0.610." probe. Only minor denting was noted at the higher supports.

The explosive plugs that were identified to be leaking at a very low rate (approximately two drops per minute) will not be weld repaired this outage. This decision is based on personnel radiation exposure associated with performing weld repairs, the low primary-to-secondary leak rate (less than 10 gallons per day) prior to the outage, and the planned steam generator replacement scheduled in 1983.

An eddy current exam of 11 (Note: One sleeved tube was removed from service in 03/82.) tubes sleeved during the 10/81 refueling outage was also performed this outage. An eddy current signal was identified in the hot leg sleeve in tube R28C58. This signal could be an indication of a deposit on the ID of the sleeve wall. Evaluations of this signal, including reprobing of this sleeved tube, have been initiated. The results of these evaluations and a description of corrective action, if necessary, will be reported in a supplement to this LER.

The roll transition and brazed areas of the tube sleeves were also inspected using the same eddy current parameters as used during the baseline inspection of 10/81. The data resulting from this inspection was compared to the baseline data, and no noticable changes in the eddy current signals were identified.

In addition to the sleeve inspections, the hot leg tubes containing sleeves were inspected through the sixth support from the hot leg side. This inspection was performed with normal eddy current parameters and a 0.650" diameter probe. No indications were identified.

To minimize the rate of corrosion, the Unit 1 primary system will be returned to power at a reduced hot leg temperature of 557*F.

The NRC Resident Inspector had been notified of these findings. This event is reportable in accordance with Technical Specification 15.6.9.A.3 and is similar to others.