ML18153A129

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LER 97-006-00:on 970320,loss of Refueling Integrity Due to Inadequate Containment Closure Process & Verification.Fuel Movement Stopped IAW Action Statement Requirements of TS 3.10.B.W/970418 Ltr
ML18153A129
Person / Time
Site: Surry Dominion icon.png
Issue date: 04/18/1997
From: Christian D
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
97-218, LER-97-006, LER-97-6, NUDOCS 9704230266
Download: ML18153A129 (6)
v • d • e


Text

10CFR50.73 Virginia Electric and Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 April 18, 1997 U.S. Nuclear Regulatory Commission Serial No.: 97-218 Document Control Desk SPS: mdk Washington, D. C. 20555 Docket No.: 50-280 License No.: DPR-32

Dear Sirs:

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

REPORT NUMBER 50-280/97-006-00 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.

~lyyours, JCC/_ _

D. A. Christian Station Manager Enclosure Commitments contained in this letter: None.

pc: Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 R. A. Musser NRG Senior Resident Inspector Surry Power Station 9704230266 970418 11111111111111m 1111111m 11111 ~" 1111 PDR ADOCK 05000280

  • I II I 3 4 Z

NRG FORM366 (4-95)

. NUCLEAR REGULATORY COMMISSION e APPROVED BY 0MB NO. 3150-0104 EXPIRES 4/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVENT REPORT (LER) LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33),

U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC (See reverse for required number of digits/characters for each block) 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE(3)

SURRY POWER STATION , Unit 1 05000 - 280 1 OF5 I TITLE (4)

Loss of Refueling lnteQritv Due to Inadequate Containment Closure Process.

EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED 8)

SEQUENTIAL REVISION FACILITY NAME DOCUMENT NUMBER MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 05000-FACILITY NAME DOCUMENT NUMBER 03 20 97 97 -- 006 -- 00 04 18 97 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)

LEVEL (10) 000  % 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) 50.7;3(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c:)(1) 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 NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)

NAME D. A. Christian, Station Manager I {;;;)N;~u;:;R~~~e Area Code)

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TONPRDS NO SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR I YES (If yes, complete EXPECTED SUBMISSION DATE).

IX I NO SUBMISSION DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

At 1206 hours0.014 days <br />0.335 hours <br />0.00199 weeks <br />4.58883e-4 months <br /> on March 20, 1997, with Unit 1 in refueling shutdown and fuel movement underway, and Unit 2 at 100% power, a management walkdown of selected refueling containment integrity penetrations questioned the adequacy of the main steam safety valve inlet pipe flange cover installation. Investigation by a maintenance team determined that the flange cover on main steam safety valve, 1-MS-SV-103C, had a gap of approximately one-eighth inch. Technical Specification 3.1 O.A.1, regarding refueling operations, requires penetrations which provide a direct path from containment atmosphere to the outside atmosphere to be closed by a .valve, blind flange, or equivalent. Since refueling containment integrity was not satisfied, fuel movement was stopped in.

accordance with the action requirements of Technical Specification 3.10.8. A containment integrity verification plan was developed and implemented. Additional containment integrity issues were identified, corrected and evaluated. .The health and safety of the public were not affected by this*

event. A fuel handling accident or radiological release did not occur. This event was caused by inadequate refueling containment closure process and verification. A Root Cause Evaluation is underway. This report is being made pursuant to 10CFR50. 73(a)(2)(i)(B) for any operation or condition prohibited by Technical Specifications.

NRC FORM 366 (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

.. LICENSEEEVENTREPORT (LEffl TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER Surrv Unit 1 05000-280 97 --006 -- 00 2 OF 5 TEXT (If more space is required, use additional copies of NRG Form 366A) (17)

1.0 DESCRIPTION

OF THE EVENT Technical Specification 3.1 O.A.1 requires refueling .containment integrity during refueling operations. This specification requires penetrations [EIIS: NH, PEN] which provide a direct path from containment atmosphere to the outside atmosphere to be closed by a valve, blind flange, or equivalent. '

During performance of scheduled Unit 1 refueling activities, main steam safety valves [EIIS: SB, RV] were removed for setpoint testing and maintenance. In accordance with the maintenance procedure that removed the valves, a blank flange cover was installed on the safety valve inlet pipe flange for foreign material exclusion purposes. The old gasket remained installed to provide a seal between the piping flange and blank flange cover. Subsequently, refueling*

containment integrity, required by Technical Specification 3.1 O.A.1, was established in accordance with the refueling containment integrity procedure. The refueling containment integrity procedure provided the requirements for establishing and verifying that containment penetrations [EIIS: NH, PEN] were properly sealed. At 1206 hours0.014 days <br />0.335 hours <br />0.00199 weeks <br />4.58883e-4 months <br /> on March 29, 1997, during a management walkdown of selected refueling containment integrity penetrations, the adequacy of the main steam safety valve blank flange covers was questioned. Further investigation by a maintenance team determined that the flange cover on the "C" main steam safety valve, 1-MS-SV-103C, had a gap of approximately one-eighth inch. The gap between the "C" main steam safety valve flange and cover did not satisfy the procedural requirements for refueling containment integrity.

Fuel off-load had commenced at 0856. hours on March 20, 1997, and fuel off-load was stopped in accordance with the action requirements of Technical Specification 3.1 O.B at 1206 hours0.014 days <br />0.335 hours <br />0.00199 weeks <br />4.58883e-4 months <br /> on March 20, 1997. The main steam safety valve inlet pipe flange cover was subsequently installed correctly and the closure bolts for all flange covers tightened. Fuel off-load recommenced at 1542 hours0.0178 days <br />0.428 hours <br />0.00255 weeks <br />5.86731e-4 months <br /> on March 20, 1997.

At 1054 hours0.0122 days <br />0.293 hours <br />0.00174 weeks <br />4.01047e-4 months <br /> on March 21, 1997, an independent inspection by a nuclear oversight specialist identified a gap in the flange cover installed over the equipment hatch emergency escape trunk containment penetration [EIIS: NH, AL, PEN]. Further investigation identified the raised surface of a seam weld on the flange cover caused a visible gap between the mating surfaces of the cover and the flange. Since the gap between the equipment hatch cover and flange allowed a direct path from containment atmosphere to the outside atmosphere, refueling containment integrity in accordance with Technical Specification 3.1 O.A.1 was not satisfied.

A containment integrity verification plan was developed and implemented to verify proper installation of refueling containment integrity covers and to verify actual valve position in accordance with tag-out boundaries.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

  • , LICENSEE EVENT REPORT (LE~

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER Surrv Unit 1 05000-280 97 --006 -- 00 3 OF 5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

At approximately 1305 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.965525e-4 months <br /> on March 21, 1997, during implementation of the containment integrity verification plan, it was determined that main steam trip valve [EIIS: SB, ISV], 1-MS-TV-101 C, was no longer fully closed. Instrument air leaking past an isolation valve [EIIS: LD, ISV]

to the air actuation cylinder was causing the trip valve actuator to hold the valve disc slightly off its seat. At 2046 hours0.0237 days <br />0.568 hours <br />0.00338 weeks <br />7.78503e-4 months <br /> on March 21, 1997, also during implementation of the containment integrity verification plan, minor leakage was identified at the equipment hatch emergency escape trunk cover penetrations. The valve and penetrations had been previously verified to be properly closed prior to fuel off-load activities.

Fuel off-load had been stopped at 1054 hours0.0122 days <br />0.293 hours <br />0.00174 weeks <br />4.01047e-4 months <br /> on March 21, 1997, upon discovery of the equipment hatch emergency escape trunk flange cover gap and did.not recommence until 2353 hours0.0272 days <br />0.654 hours <br />0.00389 weeks <br />8.953165e-4 months <br /> on March 21, 1997, after the containment integrity verification plan issues were resolved.

This report is being made pursuant to 10CFR50.73(a)(2)(i)(B) for any operation or condition prohibited by Technical Specifications, since the gap in the main steam safety valve inlet pipe flange cover and the gap in the equipment hatch emergency escape trunk cover did not satisfy Technical Specification 3.1 O.A.1 requirements for refueling containment integrity.

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS The health and safety of the public were not affected by this event. A fuel handling accident or radiological release did not occur. There are no safety consequences or implications associated with this event. The requirements of 10 CFR 100 would not have been exceeded during this event, even if a postulated fuel handling accident had occurred with the containment integrity conditions identified.

For the conditions involving the leakage paths from containment atmosphere through the main steam safety valve flange cover and the main steam trip valve, no leakage of any safety significance would have occurred due to the design configuration and existing condition of the steam generators [EIIS: SJ, HX]. The feedwater ring [EIIS: SJ, PSX] inside the steam generator is maintained full of water by a loop seal design. This feedwater ring water seal would block any direct release path to the outside atmosphere. At the time of the event, the "C" steam generator feedwater ring was filled with water. There was no direct release path from containment atmosphere to the outside atmosphere. The conditions of the event involving the main steam safety valve flange cover and the main steam trip valve were bounded by the initial conditions assumed in the UFSAR Fuel Handling Accident Analyses. No increase in accident consequences would have occurred as a result of the conditions identified above.

For those conditions involving the equipment hatch emergency escape tr.unk cover and its penetrations, an engineering evaluation determined that the observed breaches in the equipment hatch cover did not represent a compromise in refueling containment integrity which would have had a discernible impact on offsite releases in the event of a Fuel Handling Accident. There were no credible mechanisms present to create a significant pressure NRC FORM 366A (4-95)

NRG FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

' TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER Surry Unit 1 05000-280 97 --006 -- 00 40F 5 -

TEXT (If more space is required, use additional copies of NRG Form 366A) (17) differential between the containment atmosphere and the outside atmosphere. If a postulated Fuel Handling Accident were to occur with the equipment hatch conditions identified, any release would have been by diffusion and would have been negligible. Thus for the observed conditions, the limiting case results presented in the UFSAR Fuel Handling Accident Analyses remained bounding.

  • 3.0 CAUSE OF THE EVENT This event was caused by inadequate refueling containment closure process and verification. A Root Cause Evaluation is underway.

4.0 IMMEDIATE CORRECTIVE ACTION Upon discovery that refueling containment integrity was not satisfied, fuel movement was stopped in accordance with the action statement requirements of Technical Specification 3.10.8.

The main steam safety valve flange cover was installed correctly and the closure bolts tightened.

At 1527 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.810235e-4 months <br /> on March 20, 1997, the main steam safety valve flange covers were verified to be installed correctly.

A refueling containment integrity verification plan was also developed and implemented. The refueling containment integrity verification plan included:

  • Verification of proper installation of refueling containment integrity covers by the refueling Senior Reactor Operator and Mechanical Maintenance personnel.
  • Verification of actual valve position in accordance with tag-out boundaries.
  • Review of previous events related to loss of refueling containment integrity to ensure they were factored into the integrity review.
  • Heightened sensitivity to work.orders affecting containment integrity.

5.0 ADDITIONAL CORRECTIVE ACTIONS A sealant was applied on the inside and outside of the equipment hatch emergency escape trunk flange cover. Checks for air leakage and visible gaps were completed. Proper installation of the equipment hatch emergency escape trunk flange cover and its penetrations was verified. At 2157 hours0.025 days <br />0.599 hours <br />0.00357 weeks <br />8.207385e-4 months <br /> on March 21, 1997, the required actions for equipment hatch integrity were verified to be satisfactory.

_ The equipment hatch emergency escape trunk flange cover weld bead was ground flush with the face of the flange cover to ensure a proper sealing surface existed.

ThE3 affected refueling containment integrity procedures were revised. Maintenance procedure revisions included additional details for installing the equipment hatch emergency escape trunk flange cover and the main steam safety valve inlet pipe flange covers for establishing refueling containment integrity. These instructions provide criteria for assuring that an atmospheric seal is established and verified by maintenance personnel. Also, the operations procedures were revised to verify installation was accomplished in accordance with the maintenance procedures.

NRC FORM 366A (4-95)

NRC FORM 366A (4-95) e U.S. NUCLEAR REGULATORY COMMISSION

  • LlCENSEEEVENTREPORT (LE~

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER Surrv Unit 1 05000-280 97 -~oos -- oo 5 OF 5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

The revised procedures were utilized to establish containment integrity for core reload, which commenced at 2356 hours0.0273 days <br />0.654 hours <br />0.0039 weeks <br />8.96458e-4 months <br /> on April 2, 1997.

6.0 ACTIONS TO PREVENT RECURRENCE A Root Cause Evaluation is being performed. Approved recommendations from the Root Cause Evaluation will be implemented in accordance with the Corrective Action Program. The Root Cause Evaluation will address the effectiveness of corrective actions associated with previous refueling containment integrity events.

7.0 SIMILAR EVENTS

  • LER 50-280 / 92-005-00, Loss Of Refueling Containment Integrity Due To Inadequate Procedures And Work Practices.
  • LER 50-281 / 89-001-00, Loss Of Containment Integrity During Refueling Operations Due To Loss Of Administrative Controls.
  • LER 50-280 / 83-015-00, Loss of Refueling Containment Integrity Due to Nitrogen Purge Valve Removal.
  • LER 50-281 / 81-078-00, Loss of Refueling Containment Integrity Due to Removal of Inside and Outside Trip yalves.
  • 8.0 ADDITIONAL INFORMATION Unit 2 was at 100% power during this event.

NRG FORM 366A (4-95)

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