ML18152A586

LER 87-025-00:on 871008,RCS Leakage Rate Test Results of 4.06 Gpm Found in Excess of Tech Specs 3.1.C.2 & 3.1.C.5. Caused by Inadequate Procedure & HX Bypass Valve Leakage. Valve Tightened & Procedure PT-10 revised.W/871106 Ltr

ML18152A586
Person / Time
Site:	Surry
Issue date:	11/05/1987
From:	Benson D VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
87-030, 87-30, LER-87-025, LER-87-25, NUDOCS 8711130208
Download: ML18152A586 (4)

Text

POW 28-06-01 NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION.

(9-83)

APPROVED DMB NO. 3160-0104 LICENSEE EVENT REPORT (LEA) EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) I P ,GE 131 Surrv Power Station Unit 1 o 15 Io Io I oI2I 810 1 loF OI 3 TITLE (41 Excessive Reactor Coolant Svstem Leakaqe Due To Valve Seat ' Leakaae EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR ,\f se~i~~~kAL t? ~~~~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBER(S) 1 I o o Is s 1 al 1 - o I2 I s - olo 1 I1 oIs sI 1 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: /ChllCk on* or more of th* folla"(ing/ (11)

OPERATING MODE (9)

I N 20.402(b)

- 20.405(c) 60.73(1)(2)(1¥) 73.71(b) 20.405(11(1 l(i)

- 60.38(c)(1) 60.73(1)(2)(v) 73.71(c)

POWER LEVEL ~

1101 1 I n I n 20.405(1)(1 l(ji)

- 50.38(cH21 60.73(1)(2)(vll)

- OTHER (Specify in Abstract b1/ow and in T*Kt. NRC Form IIJ\lli=

20.405(1)(1 Hiiil 50.73(1)(2)(;) . 60.73(1)(21(vlll)(A) 366A}

~

20.405(1)(1 )(Iv) 20.405(1111 )(y) - 60.73(11(21(ii) 60.7311H2Hilll LICENSEE' CONTACT FOR THIS LER 112) 60.73(1H2HvlllHBI 60.73(1)(2)(11)

NAME TELEPHONE NUMBER AREA CODE D. L. Benson, Station Manager COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

MANUFAC- MANUFAC*

CAUSE SYSTEM COMPONENT TURER TURER I I I I I I I I I I I I I I I I I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED n YES (If v*s, complete EXPECTED SUBMISSION DATE}

ABSTRACT (Limit to 1400 spaces, i.e.* approximstely fifte*n single-spac* typ*writton fin**/ 116)

SUBMISSION DATE (16)

I I I On October 8, 1987 at 1208 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.59644e-4 months <br />, with Unit 1 at 100% power, a 30 minute Reactor Coolant System (RCS) {EIIS-AB} leak rate test indicated total RCS leakage of 4.06 gpm, of which 1.13 gpm was unidentified. This exceeded Technical Specification 3.1.C.2 which limits unidentified RCS leakage to less than 1 gpm. Suspecting that the leakage originated in the letdown system {EIIS-CB}, the unit was placed on excess letdown; normal letdown was secured, and a 14 minute leak rate calculation was performed at 1443 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.490615e-4 months <br />.

The results indicated total RCS leakage of 13 gpm of which .24 gpm was unidentified. This exceeded Technical Specification 3.1.C.5 which limits total RCS leakage to less than 10 gpm. Excess letdown was then secured and normal letdown reestablished. At 1702 hours0.0197 days <br />0.473 hours <br />0.00281 weeks <br />6.47611e-4 months <br />, a 67 minute leak rate calculation showed total RCS leakage of 4.02 gpm of which .24 gpm was unidentified. Since a RCS walkdown discovered no leakage and subsequent leak rates indicated unidentified RCS leakage to be acceptable, it is suspected that the original leak rate calculation was in error.

The procedure has been revised to more a.ccurately account for fluctuations in RCS temperature during leak rate calculations. It was subsequently determined that the excess letdown heat exchanger bypass valve (1-RC-107) {EIIS-V} was leaking past the seat. The valve was tightened and the seat leakage was reduced. i1/

8711130208 871105 PDR ADOCK 05000280 1~ \\ '\

NRC Form 366 S PDR (9-831

POW 28-06-01 NRC Farm 3118A e e U.S. NUCLEAR REGULATORY COMMISSION 19-831 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION APPROVED 0MB NO. 3150-0104

-i. EXPIRES: 8/31 /BB FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 161 PAGE 131 Surry Power Station, Unit 1 o 15 Io Io Io I 218 I o RI 7 - o I JI i:; - n In n I '> OF n I ':2 TEXT (If - - Is ,equlted, u* ~ NRC Form .IBA'*J 1171 1.0 Description of the Event On October 8, 1987 at 1208 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.59644e-4 months <br />, with Unit 1 at 100% power, the results of a 30 minute Reactor Coolant System (RCS) {EIIS-AB} leak rate test indicated total RCS leakage of 4.06 gpm, of which 1.13 gpm was unidentified. This exceeded Technical Specification 3.1.0.2 which limits unidentified RCS leakage to less than 1 gpm.

At 1235 hours0.0143 days <br />0.343 hours <br />0.00204 weeks <br />4.699175e-4 months <br />, a containment walkdown was conducted in an effort to determine the source of the leakage, however, none was identified. It was then suspected that the leakage originated in the letdown system {EIIS-CB}, and the unit was placed on excess letdown; normal letdown was secured, and a 14 minute leak rate calculation was performed at 1443 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.490615e-4 months <br />. The results indicated total RCS leakage of 13 gpm of which

.24 gp~ was unidentified. This exceeded Technical Specification 3.1.C.5 which limits total RCS leakage to less than 10 gpm. Excess letdown was then secured and normal letdown reestablished at 1535 hours0.0178 days <br />0.426 hours <br />0.00254 weeks <br />5.840675e-4 months <br />.

At 1702 hours0.0197 days <br />0.473 hours <br />0.00281 weeks <br />6.47611e-4 months <br />, a 67 minute leak rate calculation was completed. Results indicated total RCS leakage of 4.02 gpm of which .24 gpm was unidentified.

2.0 Safety Consequences and Implications A limited amount of leakage from the RCS is expected, but the maximum allowable values are 1 gpm from unidentified sources and 10 gpm from other than controlled sources. These values are sufficiently low to ensure early detection of unidentified or excessive leakage.

During this event, the normal charging system was in service and was able to make up for all RCS leakage.

Therefore, the he.al th and safety of the public were not affected.

NRC FORM 366A

• U.S.GPO: 1986-0*824*538/455 (9-831

• ---* -*-*--; ., ........................ : r * , - - .... _ .. .... , J . . . . . .~ - ... ~- __ ,,_ _ _ *- -:-----

POW 28-06-01 NAC Form3MA e e U.S. NUCLEAR REGULATORY COMMISSION (9-83)

LICENSEE EVENT REPORT (LERI TEXT CONTINUATION APPROVED 0MB NO. 3150-0104 EXPIRES: 8/31 /88 FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (61 PAGE (31 YEAR  ::=:,;:;:; SEQUENTIAL  ::::::::::- REVISION

..... NUMREA .*.*.*,*,* NUMBER Surry Power Station, Unit 1 o I5 I o I o jt> 12 I 8 I O RI 7 - o I 2 Ii:; - n In n I :z OF n I:z TEXT (If,_ - ii,.,,..,.,, - ~ NRC Fonn .IIIIA'*I (171 3.0 Cause Since the containment walkdown discovered no unidentified leakage and subsequent leak rates indicated unidentified RCS leakage to be well below the Technical Specification limit, it is suspected that the original leak rate calculation was in error. The leak rate procedure (PT-10) did not adequately account for fluctuations in RCS temperature during short leak rate calculations.

It was subsequently determined that the excess letdown heat exchanger bypass valve (1-RC-107) {EIIS-V}

was leaking past the seat. This was the cause of the excessive total leakage noted at 1443 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.490615e-4 months <br />, when the unit was on excess letdown.

4.0 Immediate Corrective Action A containment walkdown was conducted in an effort to identify any RCS leakage. None was found.

Additional leak rate tests were performed. Excess letdown was secured and normal letdown reestablished.

5.0 Additional Corrective Action PT-10, the Reactor Coolant Leakage procedure, has been revised to more accurately account for fluctuations in RCS temperature during leak rate calculations.

The excess letdown heat exchanger bypass valve was tightened and the seat leakage was reduced.

6.0 Action Taken to Prevent Recurrence None required.

7.0 Similar Events None.

8.0 Manufacturer/Model Number N/A NRC FORM 366A (9-831

• U.S.GPO: 1986*0-824-538/455

e VIRGINIA ELECTRIC AND POWER COMPANY Surry Power Station P. 0. Box 316 Surry, Virginia 23883 November 6, 1987 U.S. Nuclear Regulatory Commission Serial No.: 87-030 Document Control Desk Docket No.: 50-280 016 Phillips Building Licensee No. : DPR-32 Washington, D.C. 20555 Gentlemen:

Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report for Surry Unit 1.

REPORT NUMBER 87-025-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control.

Very truly yours, David L. Benson Station Manager Enclosure cc: Dr. J. Nelson Grace Regional Administrator Suite 2900 101 Marietta Street, :NW Atlanta, Georgia 30323