05000280/LER-1997-001, :on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation Circuitry

:on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation Circuitry

ML18153A142
Person / Time
Site:	Surry
Issue date:	06/10/1997
From:	Christian D VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
97-101A, LER-97-001, LER-97-1, NUDOCS 9706180217
Download: ML18153A142 (6)

LER-1997-001, on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation Circuitry

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(iv), System Actuation
2801997001R00 - NRC Website

text

JUNE 10, 1997 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555

Dear Sirs:

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

97-101 A SPS:mdk Docket No.: 50-280 License No.: DPR-32 Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following updated Licensee Event Report applicable to Surry Power Station Unit 1.

REPORT NUMBER 50-280/97-001-01 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: None.

pc: US Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303 R. A. Musser NRC Senior Resident Inspector Surry Power Station 9706180217 970610 PDR ADOCK 05000280 S

PDR A Christian.

• ,~

l ~i n Manager I llllll f llll f lll(IIIIJ Jilli! JJII I/JJ /IIJ C

6

NRG FORM 366 (4-95)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

(See reverse for required number of digits/characters for each block)

FACILITY NAME (1)

SURRY POWER STATION, Unit 1 TITLE (4)

Shutdown Due to Steam Drain Line Weld Leak EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

MONTH DAY YEAR 01 23 97 YEAR 97 SEQUENTIAL NUMBER

-- 001 --

REVISION NUMBER 01 MONTH 06 DAY YEAR 10 91 FACILITY NAME FACILITY NAME 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 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 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104).

OFFICE OF MANAGEMENT AND

BUDGET, WASHINGTON. DC 20503.

DOCKET NUMBER (2)

PAGE (3) 05000 - 280 1 OF5 OTHER FACILITIES INVOLVED 8)

DOCUMENT NUMBER 05000-DOCUMENT NUMBER 05000-THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)

OPERATING MODE (9) n 20.2201 (b) 20.2203(a)(2}(v) x 50.73(a)(2}(i) 50.73(a)(2)(viii)

POWER LEVEL (10) 100 %

20.2203(a)(1) 20.2203(a)(3}(i) 50.73(a)(2)(ii) 50.73(a)(2}(x) 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) 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 LER (12 NAME 1

TELEPHONE NUMBER (Include Area Code)

D. A. Christian, Station Manager (757) 365-2000 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS X

IG DET x999 y

SUPPLEMENTAL REPORT EXPECTED (14)

I YES I X I NO (If yes, complete EXPECTED SUBMISSION DATE).

EXPECTED SUBMISSION DATE MONTH DAY TONPRDS YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced tyoewritten lines) (16)

At 1850 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.03925e-4 months <br /> on January 23, 1997, with Unit r*and Unit 2 both at 100% power, a 'Service Building Operator performing routine activities in the Unit 1 Main Steam Valve House discovered a small steam leak in the steam drain piping on the " B" main steam line. Unit 1 was shutdown in accordance with the. requirements of Technical Specification 4.15.C.1.

While performing normal shutdown procedures, both source range nuclear instruments failed to indicate source range counts, and operator action to manually trip the turbine from the main control room failed to result in a turbine trip.

The turbine was tripped locally at the governor pedestal using the manual trip lever. While shutdown, both source range detectors were replaced, tested and the nuclear instruments were returned to service. The turbine trip actuation circuitry was inspected and tested satisfactorily. The leak was repaired, inspected, and tested prior to returning the Unit to power operation. The health and safety of the public were not affected by this event. A radiological release did not occur. This event was caused by a pin hole leak in a 1 1/2 inch diameter weld. This report is being made pursuant to 1 OCFR50. 73(a)(2)(i)(A), for any nuclear plant shutdown required by the plant's Technical Soecifications.

NRG FORM 366 (4-95)

i (4-95)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

Surry Unit 1 TEXT (If more space is required, use additional copies of NRG Form 366A) (17) 1.0 DESCRIPTION OF THE EVENT DOCKET 05000-280 LEA NUMBER (6)

PAGE (3)

YEAR I SEQUENTIAL I REVISION NUMBER

• NUMBER 97

--001 --

01 2 OF 5 Surry Technical Specification 4.15 contains an Augmented Inspection Program requirement that all welds in the Main Steam Valve House receive a visual inspection of the surface of the insulation at all weld locations on a weekly basis for detection of leaks. Any detected leaks shall be investigated and evaluated. If the leakage is caused by a through-wall flaw, either the* plant shall be shutdown, or the leaking piping isolated. Repairs shall be performed prior to return of the line to service.

At approximately 1850 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.03925e-4 months <br /> on January 23, 1997, a Service Building Operator performing routine duties in the Unit 1 Main Steam Valve House heard a small leak in the vicinity of the steam drain piping on the "B" main steam [EIIS:SB] line.

Following further investigation, the Service Building Operator notified the Unit 1 Control Room Operator who informed the Senior Reactor Operator of a steam leak. Maintenance and engineering personnel were assigned to remove the piping insulation and investigate the leak. The insulation was removed by 2225 hours0.0258 days <br />0.618 hours <br />0.00368 weeks <br />8.466125e-4 months <br /> on January 23, 1997 and Engineering personnel identified a pin hole leak approximately 1/32 inch in diameter through a weld in main steam drain line, 1 1/2-SHPD-8-601. Since the steam leak was caused by a through-wall flaw in a weld location that was unisolable, Technical Specification 4.15.C.1 required the plant to be shutdown and repairs performed prior to returning the line to service. At 0124 hours0.00144 days <br />0.0344 hours <br />2.050265e-4 weeks <br />4.7182e-5 months <br /> on January 24, 1997, a Unit 1 shutdown was commenced. At 0143 hours0.00166 days <br />0.0397 hours <br />2.364418e-4 weeks <br />5.44115e-5 months <br /> on January 24, 1997, a one-hour report was made to the NRG in accordance with 10 CFR 50.72(b)(1 )(i)(A) for initiation of any plant shutdown required by Technical Specifications. The Unit was shutdown by 0817 hours0.00946 days <br />0.227 hours <br />0.00135 weeks <br />3.108685e-4 months <br /> on January 24, 1997. A flaw characterization and weld repair were initiated. The failure mechanism was slag entrapment ranging in size from 1/32 to 3/16 inches, located in the weldment.

The flaw was removed by grinding and the weld was repaired.

A visual inspection and pressure test were performed and the line was returned to service.

This event is reportable in accordance with 10 CFR 50. 73(a)(2)(i)(A) for any plant shutdown required by Technical Specifications.

While ramping the Unit off-line in accordance with normal shutdown procedures the following equipment failures occurred. At.0718 hours0.00831 days <br />0.199 hours <br />0.00119 weeks <br />2.73199e-4 months <br /> on January 24, 1997, when reactor power had been decreased and stabilized at 2 percent, efforts to manually trip the turbine [EIIS:TA] from the Main Control Room using the turbine trip push-buttons [EIIS:IT] in accordance with normal shutdown procedures were unsuccessful. The turbine was tripped locally, at the turbine, using the manual trip lever at 0721 hours0.00834 days <br />0.2 hours <br />0.00119 weeks <br />2.743405e-4 months <br />, and the main generator [EIIS:TB] output breakers were opened. Also, once reactor power was below the point of adding heat and the reactor was tripped in accordance with normal shutdown procedures with the control rods fully inserted and reactor trip breakers open, both source range nuclear instruments [EIIS:IG,JI] energized as expected and then failed to indicate source range counts.

i (4-95)

FACILITY NAME (1)

Surry Unit 1 LICENSEE EVENT REPORT {LER)

TEXT CONTINUATION DOCKET U.S. NUCLEAR REGULATORY COMMISSION LER NUMBER (6)

PAGE (3)

YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 05000-280 97

--001 --

01 3 OF 5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

At 0745 hours0.00862 days <br />0.207 hours <br />0.00123 weeks <br />2.834725e-4 months <br /> with both source range nuclear instruments declared inoperable, Abnormal Procedure, 0-AP-4.00, was implemented.

The plant was stabilized in the hot shutdown condition with the control rods fully inserted, the reactor trip breakers open, and adequate shutdown margin verified consistent with Action 5 of Technical Specification Table 3.7-1, Item 4.b.

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS

There are no safety consequences or implications associated with this event.

Flaw characterization and examinations by materials engineering personnel determined the pin hole leak in the weld was an isolated occurrence caused by slag inclusion during original construction which allowed the flaw to slowly propagate to the weld surface after 25 years of service.

The failure of the turbine to trip manually using the Main Control Room push-buttons was a previously analyzed event for which procedures and training were in place to locally trip the turbine using the manual trip lever. Normal shutdown procedures require that reactor power be reduced to less than 2 percent and stabilized prior to tripping the turbine. These procedures allow use of either the Main Control Room push-buttons or the local trip lever to execute a turbine trip and were properly executed during the event. There is no safety significance related to this portion of the event since reactor power had been reduced to 2 percent and stabilized prior to tripping the turbine.

• The reactor trip function associated with a turbine trip is allowed to be blocked below 10 percent reactor power. Subsequent testing verified that the automatic turbine trip from a reactor trip signal remained operable during this event.

Both source range nuclear instruments energized as required during the shutdown evolution, indicated source range counts, and then fail$-d to properly indicate source range counts after less than 1 O minutes of operation.

Failure of the source range nuclear instruments had been*

previously evaluated.

Procedures and training were in place should such an event occur.

Normal shutdown procedures require that reactor power be reduced below the point of adding heat and stabilized before tripping the reactor and energizing the source range detectors. Upon discovery that the source range nuclear instruments failed, appropriate procedures were properly executed ensuring the plant remained stable in the safe shutdown condition, with all controls rods fully inserted, reactor trip breakers open, and adequate shutdown margin in place in accordance with Action 5 of Technical Specification Table 3.7-1, Item 4.b.

At no time were the health and safety of the public or plant personnel affected by this event.

3.0 CAUSE OF THE EVENT

Flaw characterization and examinations by materials engineering personnel determined that the pin hole leak in the weld was an isolated occurrence caused by slag inclusion during original construction which allowed a small flaw to slowly propagate to the weld surface after 25 years of service.-(4-95)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LE~

TEXT CONTINUATION FACILITY NAME (1)

Surry Unit 1 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

DOCKET 05000 -280 LEA NUMBER (6)

PAGE (3)

YEAR I SEQUENTIAL l REVISION NUMBER NUMBER 97

--001 --

01 4 OF 5 The unexpected failure of the turbine to trip manually using the Main Control Room push-buttons

• was initially believed to be due to a sluggish auto stop drain valve.

Subsequently it was determined to be due to binding within the turbine protective trip block assembly attributable to the lack of turbine protective trip block testing.

The unexpected failure of the source range detectors was initially believed to have been caused by aging. Subsequently this cause was confirmed.

4.0 IMMEDIATE CORRECTIVE ACTION

Upon discovery of the steam leak in the Main Steam Valve House the Service Building Operator immediately notified the Unit 1 Control Room Operator. The Unit 1 Senior Reactor Operator notified maintenance personnel and engineering personnel to investigate the leak, and management was notified. Following removal of the piping insulation by maintenance personnel, engineering personnel investigated the leak source and identified the leak to be a pin hole through the wall of a weld in a 1 1/2 inch drain line that could not be isolated. Management was notified and the Shift Supervisor invoked the requirements of Technical Specification 4.15.C.1 and a Unit shutdown commenced.

Upon failure of the turbine to trip when the manual turbine trip push-buttons in the Main Control Room were depressed, the local turbine trip lever on the governor end pedestal was used to trip the turbine. The normal procedure for "Turbine - Generator Shutdown" allows use of either trip mechanism to execute a turbine trip. Following execution of the turbine trip using the turbine trip lever, the main steam stop valves, control valves, reheat stop valves and intercept valves were verified closed; the auto stop oil header was verified to be depressurized; the extraction steam non-return valves were verified shut; and turbine shaft speed was verified to be decreasing.

Upon failure of the source range nuclear instruments to continue, indicating source range counts, Abnormal Procedure, O-AP-4.00, was initiated and the plant was stabilized in the safe shutdown condition. All control rods were verified to be fully inserted, the dilute function to the blender was l maintained under administrative control when not secured, the control rod drive motor /

generator supply breakers were racked to the out position preventing any possible control rod movement, adequate shutdown margin was verified, and positive reactivity additions were prohibited.

The gammametric nuclear instruments were verified operable and used as an alternate indication of source range counts. The requirements of Technical Specifications Table

3. 7-1, Item 4, Action 5 were initiated and management was notified.

5.0 ADDITIONAL CORRECTIVE ACTIONS

The flaw in the weld was excavated by grinding out the entire portion of the weld encompassing the flaw and performing a weld repair. A visual examination and pressure test were performed to ensure a quality repair was made.

• The insulation was removed from similar welds on the remaining Unit 1 steam drain lines and a visual inspection was performed with no indication of Ii

\\ (4-95)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

Surry Unit 1 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

DOCKET 05000-280 LEA NUMBER (6)

PAGE (3)

YEAR I SEQUENTIAL l REVISION NUMBER NUMBER 97

--001 --

01 5 OF 5 leakage or welding flaws. The piping in the Unit 2 Main Steam Valve House was walked down in accordance with surveillance procedures to ensure a similar leak did not exist.

The turbine trip circuits were tested and verified operational. The turbine control block latch mechanism and suspected sluggish drain valve were inspected and tested to ensure proper operation. A turbine trip functional test was performed to verify proper operation of the manual turbine trip push-buttons and the turbine trip circuit.

The source range nuclear instrument vendor was contacted and a representative was brought on site to assist in the troubleshooting and repair. The source range nuclear instrument, N-32, detector was replaced, tested and returned to service on January 31, 1997. The source range nuclear instrument, N-31, detector was replaced, tested and returned to service on February 1, 1997.

6.0 ACTIONS TO PREVENT RECURRENCE Related to the weld leak, the Augmented Inspection Program requirements will be conducted in accordance with Technical Specification 4.15.

A Root Cause Evaluation, as well as a component functional failure evaluation, required by our Maintenance Rule Program in accordance with 10 CFR 50.65, was performed to investigate the failure of the turbine to trip manually using the Main Control Room push-buttons. This failure was determined to be a Maintenance Rule Functional Failure due to binding within the turbine protective trip block assembly attributable to the lack of turbine protective trip block testing.

Approved recommendations from Root Cause Evaluations are implemented in accordance with the Corrective Action Program.

To investigate the source range nuclear instrument failures, a Root Cause Evaluation was performed with a multidisciplined root cause evaluation team consisting of craft, technical, and engineering personnel assigned to investigate the root cause and recommend corrective actions.

The Root Cause Evaluation, as well as a Maintenance Rule component functional failure evaluation, has been completed.

These failures were determined to be a Maintenance Preventable Functional Failure due to component aging. Approved recommendations from Root Cause Evaluations are implemented in accordance with the Corrective Action Program.

7.0 SIMILAR EVENTS

There were no similar events identified.

8.0 ADDITIONAL INFORMATION

The Source Range Nuclear Instrument detectors are Westinghouse part number WL-24158, serial number 912101.