05000280/LER-2000-001, Containment Isolation Valve Found with Unacceptable Leakage During Maintenance

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Containment Isolation Valve Found with Unacceptable Leakage During Maintenance
ML19105A834
Person / Time
Site: Surry  Dominion icon.png
Issue date: 02/25/2000
From: Grecheck E
Virginia Electric & Power Co (VEPCO)
To:
Office of Nuclear Reactor Regulation
References
LER 2000-001-00
Download: ML19105A834 (8)
v • d • e


LER-2000-001, Containment Isolation Valve Found with Unacceptable Leakage During Maintenance
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)(ii)
10 CFR 50.73(a)(2)(iv), System Actuation
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2802000001R00 - NRC Website
v • d • e

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Subject:

Surry,Unit 2, Licensee Event Report 50-281/2000-001-00 regarding containment isolation valve found with unacceptable leakage during maintenance.

  • 1 lit,,

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D istri 96. txt Body:

ADAMS DISTRIBUTION NOTIFICATION.

Electronic Recipients can RIGHT CLICK and.OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003691561.

IE22 - 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

Docket: 05000281 Page 2

February 25, 2000 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555-0001

Dear Sirs:

e 10CFR50.73 Virginia Electric And Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 Serial No.:

00-082 SPS: BAG Docket No.: 50-281 License No.: DPR-37 Pursuant to 1 OCFR50.73, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to Surry Power Station Unit 2.

Report No. 50-281/2000-001-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to the Management Safety Review Committee for its review.

Enclosure Commitments contained in this letter:

Very truly yours, b~~

E. S. Grecheck, Site Vice President Surry Power Station

1. Upon completion of the root cause evaluation, the approved recommendations will be implemented through the Corrective Action Program.

cc:

U. S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth Street, SW, Suite 23 TBS Atlanta, Georgia 30303-8931 Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station

e e

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (6-1998)

EXPIRES 06/30/2001 Eslimaled,.the NAC may not conduct or sponsor, and a person is not required lo respond to. the information collection.

FACILITY NAME (1)

DOCKET NUMBER (2)

PAGE (3)

SURRY POWER STATION, Unit 2 05000 -281 1 OF4 TITLE (4)

Containment Isolation Valve Found with Unacceptable Leakage During Maintenance I EVENT DATE (5) I LER NUMBER (6)

EPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

SEQUENTIAL REVISION FACILITY NAME DOCUMENT NUMBER MONTH DAY YEAR YEAR MONTH DAY YEAR 05000-NUMBER NUMBER 01 27 00 00 01 00 FACILITY NAME DOCUMENT NUMBER 05000-OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)

MODE (9)

N 20.2201(b) 20.2203(a)(2)(v)

X 50.73(a)(2)(i)

50. 73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii)

50. 73(a)(2)(x)

EVEL (10) 100%

20.2203(a)(2)(i) 20.2203(a)(3)(ii)

50. 73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER 20.2203(a)(2)(iii) 50.36(c)(1)

X 50. 73(a)(2)(v)

Specify in Abstract below 20.2203(a)(2)(iv) 50.36(c)(2) 50,73(a)(2)(vii) or in NRG Form 366A, LICENSEE CONTACT FOR THIS LEA 12 NAME I (;;;)N;;;:;~;lt hea Code)

E. S. Grecheck, Site Vice President COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE TO EPIX TOEPIX X

JM ISV Crosby y

SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR I YES IX I NO SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE),

DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On January 27, 2000, a review of testing results indicated that during the replacement of 2-DA-TV-200A, the leakage through 2-DA-TV-200B was in excess of the allowable Type "C" limits. On January 25, 2000, with Unit 2 at 100% reactor power, 2-DA-TV-200A was replaced in less than one hour. The excessive leakage of 2-DA-TV-200B was not discovered until performing the post maintenance testing of 2-DA-TV-200A. Once this condition was identified, appropriate actions were taken in accordance with Technical Specifications. 2-DA-TV-200B was replaced, satisfactorily tested, and returned to service.

Based upon preliminary indications, it is believed that 2-DA-TV-200B failed because dirt and debris became impacted between the ball and valve seat causing excessive wear. A root cause evaluation has been initiated and approved recommendations will be implemented through the Corrective Action Program.

No conditions adverse to safety resulted from this event and the health and safety of the public were not affected. A 4-hour report was made on January 27, 2000 pursuant to 1 OCFR50.72(b)(2)(iii)(C). This event is being reported pursuant to 1 OCFR50.73(a}(2)(v)(C) and 1 OCFR50.73(a}(2)(i)(B).

e e U.S. NUCLEAR REGULATORY COMMISSION

  • 1 (6-1998)

FACILITY NAME (1)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET Surry Power Station, Unit 2 05000 - 281 YEAR 2000 LEA NUMBER (6)

I SEQUENTIAL I REVISION NUMBER NUMBER

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00 PAGE (3) 2 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 1.0 DESCRIPTION OF THE EVENT On January 20, 2000, a Unit 2 inside containment isolation valve, 2-DA-TV-200A.

[EII.S-JM,ISV], was cycled and the stroke time was measured to comply with the quarterly testing requirements of the lnservice Testing (1ST) Program. At 1339 hours0.0155 days <br />0.372 hours <br />0.00221 weeks <br />5.094895e-4 months <br />, the stroke time for 2-DA-TV-200A was measured at 2.59 seconds, exceeding the 1ST maximum stroke time of 2.0 seconds. 2-DA-TV-200A was declared inoperable. At 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />, the corresponding outside containment isolation valve, 2-DA-TV-200B [EIIS-JM,ISV], was closed and.

deactivated in accordance with the requirements of Technical Specification (TS) 3.8.C.1.b.

On January 25, 2000, at 1006 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.82783e-4 months <br />, with Unit 2 at 100% reactor power, a team entered containment to replace 2-DA-TV-200A. The valve was replaced and at 1055 hours0.0122 days <br />0.293 hours <br />0.00174 weeks <br />4.014275e-4 months <br />, 2-DA-TV-200A was stroked satisfactorily. 2-DA-TV-200Awas removed from the system for less than 49 minutes.

To complete post maintenance testing (PMT) of 2-DA-TV-200A, a 1 OCFR50 Appendix J Type "C" containment leak rate test was performed on January 25, 2000, at 1549 hours0.0179 days <br />0.43 hours <br />0.00256 weeks <br />5.893945e-4 months <br />.

Test results indicated that leakage was greater than 25 standard cubic feet per hour (scfh).

However, because the test air was supplied between 2-DA-TV-200A and 2-DA-TV-200B, it could not be determined which valve was leaking. The leakage was assumed to be through 2-DA-TV-200A based on a history of successful as-found Type "C" testing on 2-DA-TV-200B and the possibility that 2-DA-TV-200A became misadjusted during installation.

On January 26, 2000, the valve alignment used during the leak rate test was changed to verify that the excessive leakage was through 2-DA-TV-200A. At 1338 hours0.0155 days <br />0.372 hours <br />0.00221 weeks <br />5.09109e-4 months <br />, it was determined that the leakage through 2-DA-TV-200A was less than 0.07 scfh. The valve alignment was then changed to allow leak testing of 2-DA-TV-200B and, at 1353 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.148165e-4 months <br />, leakage through 2-DA-TV-200B was determined to be 78.86 scfh (although not considered accurate due to the line not being verified as drained). Due to the as-found leakage on 2-DA-TV-200B, a 1-hour TS action statement was entered as a conservative measure to establish containment integrity in accordance with TS 3.8.A.1 (subsequent review indicates that a 4-hour action statement in accordance with TS 3.8.C.1.b would have been acceptable). At 1405 hours0.0163 days <br />0.39 hours <br />0.00232 weeks <br />5.346025e-4 months <br />, 2-DA-TV-200A was verified closed and deactivated in accordance with the requirements of TS 3.8.C.1.b.

On January 27, 2000 at 1201 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.569805e-4 months <br />, a Type "C" containment leak rate test performed (with the line verified to be drained) on 2-DA-TV-200B determined the as-found leakage to be 182.57 scfh. The 1 OCFR50 Appendix J Type "B" & "C" total leakage acceptance criteria is 180 scfh.

A review of the subject maintenance and testing activities revealed that 2-DA-TV-200B had been inoperable during the replacement of 2-DA-TV-200A. This condition alone could have

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e e U.S. NUCLEAR REGULATORY COMMISSION

,I (6-1998)

FACILITY NAME (1)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET Surry Power Station, Unit 2 05000 - 281 YEAR LEA NUMBER (6)

I SEQUENTIAL I REVISION NUMBER NUMBER 2000

-- 001 --

00 PAGE (3) 3 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. A 4-hour non-emergency report to the NRC Operations Center was made on January 27, 2000 at 1538 hours0.0178 days <br />0.427 hours <br />0.00254 weeks <br />5.85209e-4 months <br /> pursuant to 1 OCFR50.72(b)(2)(iii)(C). This report is being made pursuant to 1 OCFR50.73(a)(2)(v)(C).

Based upon the available information regarding valve leakage and stroke times, appropriate actions were taken to ensure compliance with TSs. However, a review of the testing results revealed that 2-DA-TV-200A should have been closed and deactivated on January 25, 2000, to ensure containment integrity. Therefore, this report is also being made pursuant to 1 OCFR50.73(a)(2)(i)(B), as an operation or condition prohibited by TSs since 2-DA-TV-200A was closed but not deactivated.

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS

. The reactor containment sump pump discharges through two normally open pneumatically operated trip valves, 2-DA-TV-200A and 2-DA-TV-200B, located on each side of the containment wall. These valves function as part of the Reactor Containment Isolation System [EIIS-JM]. The valves close upon receipt of a containment isolation signal (safety injection). Containment isolation valves are tested with design pressure to ensure isolation during a design basis accident. The measured leakage under test conditions through 2-DA-TV-200B was 182.57 scfh. The measured leakage under test conditions from other pathways was 142.08 scfh. The total leakage from all pathways during the period of concern (i.e., January 25, 2000, for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when 2-DA-TV-200A was removed and 2-DA-TV-200B exceeded acceptable leakage rates) exceeded an allowable limit of 305 scfh.

Considering the conditions discovered during this event, a review of the model used to assess off-site doses from the release of radionuclides for a design basis Loss of Coolant Accident (LOCA) at Surry was performed. The review concluded that with the measured leakage from this event, the calculated off-site doses remained within the regulatory limits stated in 1 O CFR 100. The review also concluded that the potential control room doses would have remained within the limits specified by General Design Criteria 19.

The time period in which potential increased leakage through the containment sump pump discharge line existed was less than one hour (which is the allowed out of service time for containment integrity in accordance with TS 3.8.A.1.a). The probability of a LOCA occurring during this event remained small at approximately 5.0E-8.

In conclusion, this event resulted in no safety consequences or significant implications, and the health and safety of the public were not affected.

e e

" U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

FACILITY NAME (1)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET Surry Power Station, Unit 2 05000 - 281 YEAR LEA NUMBER (6)

I SEQUENTIAL I REVISION NUMBER NUMBER 2000

-- 001 --

00 PAGE (3) 4 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

3.0 CAUSE

Based upon preliminary indications, it is believed that 2-DA-TV-200B failed because dirt and debris became impacted between the ball and valve seat causing excessive wear. A root cause evaluation (RCE) has been initiated and will determine the cause of the failure.

4.0 IMMEDIATE CORRECTIVE ACTION(S)

On January 26, 2000, at 1405 hours0.0163 days <br />0.39 hours <br />0.00232 weeks <br />5.346025e-4 months <br />, after leak rate results on valve 2-DA-TV-200B were found to exceed the acceptance criteria, 2-DA-TV-200A was verified closed and deactivated in accordance with TS 3.8.C.1.b.

5.0 ADDITIONAL CORRECTIVE ACTIONS

Valve 2-DA-TV-200B was replaced, tested satisfactorily, and returned to service.

6.0 ACTIONS TO PREVENT RECURRENCE An RCE has been initiated and, upon completion, the approved recommendations from the RCE will be implemented through the Corrective Action Program.

7.0 SIMILAR EVENTS

None 8.0 MANUFACTURER/MODEL NUMBER Crosby 2 inch, 150 pound ball valve with Bettis air actuator NCB 415-SRBO, Drawing

  1. D-SC-96162

9.0 ADDITIONAL INFORMATION

Unit 1 was operating at 100% reactor power and was not affected by this event.

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