ML18054A563

LER 89-001-00:on 890127,primary Coolant Sys Leak Rate Determination Not Completed,Per Plant Tech Specs Table 4.2.2,Item 8.Caused by in-progress Plant Derate.Leak Rate Determination Completed & Evaluation ongoing.W/890227 Ltr

ML18054A563
Person / Time
Site:	Palisades
Issue date:	02/27/1989
From:	JOHNSON B D, KOZUP C S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-001-01, LER-89-1-1, NUDOCS 8903030501
Download: ML18054A563 (6)

Text

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LICENSEE EVENT REPORT (LERI IXNtU: 1n11* PM:IUT'f'

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1.ICl ... 1 CONTACT POii THIS I.Ill 1121 C S Kozup, Technical Engineer, Palisades

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• .nc.ICIHal AlllA C:COI ---71..71CM 71..71e.t ontlllll Ab1Mar ._.,.. ,,, rut. NllC "°"" >>Ml Tll.P+40flll MUMllll CAUlll SYST1111 CCWONINT x I *1 I I I I I I I I I I I I I I I I I I I I I I n Yll /It,,_, .....,._ IXl'WCTIO SIJftllSSION OA TWJ ll<,.C'nD SIJIMlllJON CATI 1191 I I I On January 27, 1989 at 0900 the primary coolant system (PCS) [AB] leak rate determination was not completed as required by Plant Technical Specification (TS) Table 4.2.2 Item 8. This determination was not completed due to an in-progress Plant derate resulting from exceeding the 0.1 gallon per minute (gpm) administrative limit for primary to secondary coolant leak.age.

The reactor was critical with the Plant operating at 45 percent of rated power at the time of the event. The required le.ak. rate determination.

completed at 1034 resulted in identified PCS leakage being calculated to be 0.0819 gpm. Due to the in-progress Plant derate, General Operation Procedure requirements could not.be met. The primary to secondary system leakage was the result of a leaking steam generator tube in the "B" steam generator.

Subsequent eddy current examinations revealed the feak.ing tube contained a

indication (approximately 57 degrees in circumference) near the top, yet within the third support plate. The crack-like indication is believed to be the result of a combination of localized stress, tube denting and support plate

"!'!*;u "::.c-:c'"*., *o*., .. *.: This event is being reported due.to the failure to _perform the required PCS leak rate determination.

Det°ails thE! steam generator tube leak are provided for information.

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,_,...'I.I (171 Description On January 27, 1989 at 0900 the primary coolant system (PCS) [AB] leak rate determination was not completed as required by Plant Technical Specification (TS) Table 4.2.2 Item 8. This determination was not completed due to an in-progress Plant derate resulting from exceeding the O.l gallon per minute (gpm) administrative limit for primary to secondary coolant leakage. The reactor was critical with the Plant operating at 45 percent of rated power at the time of the event. TS Table 4.2.2 Item 8 requires that leakage be evaluated daily. TS 4.02 allows for extending the surveillance time interval up to 25 percent of the required time. The January 26, 1989 PCS leak rate was completed at approximately 0300. Therefore, the next leak rate would be required to be completed by approximately 0900 on January 27, 1989. Due to the requirements of Plant General Operating Procedure (GOP) 13, "Primary System Leakage Calculation" and the Plant derate in-progress as detailed below, the PCS leakage calculation was not completed until 1034 on January 27, 1989. GOP 13 requires that Plant power level be stabilized and remain so during the duration of the three hour PCS leak rate determination.

• Date January 22, 1989 0445 January 22, 1989 0810-1053 January 22, 1989 1405-1500 January 24, 1989 1035-1155 . LER 89001-LIOl Description With the Plant at 100 percent of rated power, condensor off-gas radiation monitor RIA-0631 [SG;RI] count rate increased from approximately 1260 to 1700 counts per minute (cpm). Condensor of £-gas flow rate and PCS radioactivity were essentially constant.

Chemistry samples indicated no activity increases in either steam generator

[AB;SG]. Estimated primary to secondary leakage was determined to be 0.006 gpm. RIA-0631 count rate increased to 3700 cpm. Primary to secondary leakage determined to be 0.008 gpm. With the Plant at 99.9 percent of rated power, RIA-0631 :* I LICENSEE EVE.EPORT (LERJ TEXT CONTINUATIO._

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EX"llll: lr.11 Ill PALISADES NUCLEAR PLANT TIXT Ill ___ ......_ --NffC ,_...,.,IJ 1171 January 25, 1989 January 25, 1989 January 26, 1989 January 26, 1989 January 27, 1989 ""C *O"M :IMA 11-131 LER 89001-LIO l OOClllT NUllelll 121 Liii llKMUll Ill 0630 2210 0215-0410 1655-1930 0050 activity increased to 13,000 cpm. This corresponded to a primary to secondary leak rate of 0.046 gpm. Primary to secondary leakage estimated to be 0.050 gpm based on RIA-0631 activity.

Chemistry personnel reported primary to secondary leakage at 0.114 gpm based on RIA-0631 activity of 24,000 cpm. Operations personnel entered System Operating Procedure 1, "Primary Coolant System" due to the PCS system leakage exceeding the O.l gpm strative limit. TS 3.1.5.d requires Plant shutdown be initiated when primary to secondary leakage in a steam generator exceeds 0.3 gpm for any period of steady state operation greater than 24 consecutive hours. Chemistry personnel reported that the primary to secondary leak rate was 0.059 gpm and that the results reported at 2210 in January 25, 1989 were in error. Operations personnel secured from SOP 1. With the Plant operating at 100 percent of rated power, RIA-0631 activity increased to 40,000 cpm. Chemistry reported primary to secondary leakage at 0.108 gpm. Plant shutdown at a six percent per hour of rated power was ordered. With the Plant at 75 percent of rated power, PCS pressure was reduced from 2060 psi to 2000 psi.

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1131 .. DOCICIT 12t ..... *-" ,. '"°' 131 PALISADES PLANT o 1 s Io I o Io I 21 5 15 a 1 9 -o 1 o 1 i -o 1 o o 1 4 oF o 1 5 TIXT 1*--* i....-. -_,,.NltC ,_..,...,cm January 27, 1989 January 27, 1989 January 27, 1989 January 30, 1989 Cause Of The Event 0630 0711 1034 0805 Chemistry personnel reported that primary to secondary leakage has decreased to 0.017 gpm. Plant derate halted at roximately 45 percent of rated power to perform TS required PCS.leak rate mination.

PCS leak rate determination completed identified leak rate was determined to be 0.819 gpm. Plant power reduction at 12 percent per hour initiated to achieve Plant shutdown for steam generator repairs. The failure to. complete the required PCS leak rate calculation was due to the in-progress Plant reduction resulting from the primary to secondary leakage. The need to reduce Plant power, PCS pressure and the primary to secondary leak rate did not provide stable enough conditions for a three-hour period to perform the PCS leakage determination.

The primary to secondary system leakage was the result of a leaking steam generator tube in the "B" steam generator.

Subsequent eddy current examinations revealed the leaking tube contained a crack-like indication (approximately 57 degrees in circumference) near the top, yet within the third support plate. The crack-like indication is believed to be the result of a combination of localized stress, tube denting and support plate displacement.

Corrective Action The required PCS leak rate determination was completed at 1034 on January 27, 1989. An evaluation is being conducted to determine the necessity for a TS change regarding the allowance of extending PCS leak rate determinations beyond existing time intervals during Plant shutdowns.

Eddy current testing was performed utilizing the 4C4F technique for the full hot leg length of 104 steam generator tubes and utilizing the Motorized Rotating Absolute Pancaked Coil (MPRC) technique at the third support plate for 40 steam generator tubes. No additional degradation LER 89001-LIOl

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EX,.ltH: 1131119 PACIUTY NAiii 111 DOCICIT NUlimllt 121 19-331 191 'AQI IJI PALISADES NuCLEAR PLANT o 1 s 1 o 1 o 1 o 1 2 1 5 1 5 a 19 _ 9 i _ o 1 o 9 5 oF o 1 5 was noted during this testing. All active steam generator tubes surrounding-the leaking tube were plugged as a preventive measure. In addition to the above testing, 240 tubes were inspected utilizing the MRPC technique at the 13th support plate of the "B" steam generator.

The tubes inspected were in the area of previously identified (1987/1988) crack-like indications.

Crack-like indications were noted in three tubes. In conjunction with this inspection, three tubes in the 11 8 11 steam generator were deplugged and reinspected utilizing the MRPC technique in order to monitor crack propagation.

Two of these tubes exhibited no change from previous inspections, while the third indicated growth from 57 to 150 degrees in circumference.

All three tubes were replugged.

A to_tal of 238 tubes were plugged in a block. arrangement as a preventive measure to encompass apparent affected areas. This area of plugging also included all active tubes surrounding the leaker. Analysis Of The Event Technical Specifications regarding primary to secondary system leakage were not approached.

In that the FSAR safety analysis (Section 14.15) for a steam generator tube rupture assumes a double ended rupture, no safety consequence resulted from this event. This event is being reported per 10CFR50.73(a)(2)(i)(B) as an operational condition prohibited by Plant Technical Specifications regarding the failure to perform the required PCS leak rate. Additional Information For additional information regarding steam tube leaks, reference Licensee Event Reports82-012, 88-012 and 88-012-01.

For additional information regarding the failure to perform required TS surveillances, reference Licensee Event Reports83-078, 84-018,85-002, 85-004,85-012, 85-020 and 88-008. LER 89001-LIOl POWERING MICHIGAN'S PROGRESS General Offices: 1945 West Parnell Road, Jackson, Ml 49201 * (517) 788-0550 February 27, 1989 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 -LICENSE DPR-20 -PALISADES PLANT -LICENSEE EVENT REPORT 89-001 -IN-PROGRESS PLANT DERATE RESULTS IN FAILURE TO PERFORM REQUIRED SURVEILLANCE)

Licensee Event Report (LER)89-001 (In-progress Plant Derate Results in Failure to Perform Required Surveillance) is attached.

This event is reportable to the NRC per 10CFR50.73(a)(2)(i).

Brian D Johnson Staff Licensing Engineer CC Administrator, Region III, USNRC NRC Resident Inspector

-Palisades Attachment OC0289-0066-NL02