05000354/LER-2005-003, Regarding Reactor Coolant System Leak from Check Valve Position Indicator

From kanterella
(Redirected from 05000354/LER-2005-003)
Jump to navigation Jump to search
Regarding Reactor Coolant System Leak from Check Valve Position Indicator
ML052280250
Person / Time
Site: Hope Creek 
Issue date: 08/08/2005
From: Massaro M
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 05-003-00
Download: ML052280250 (4)
v • d • e


LER-2005-003, Regarding Reactor Coolant System Leak from Check Valve Position Indicator
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)
10 CFR 50.73(a)(2)(viii)(B)
3542005003R00 - NRC Website
v • d • e

text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bdidge, New Jersey 08038-0236 o PSEG Nulear LLC LR-N05-0383 AUG 0 -8 2005 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 LER 35412005-003 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 This Licensee Event Report entitled, "Reactor Coolant System Leak from Check Valve Position Indicator," is being submitted pursuant to the requirements of 16CFR50.73(a)(2)(i)(A), IOCFR50.73(a)(2)(iv)(A), and 10CFR50.73(a)(2)(ii)(A).

Sincerely, Michael J. M ssaro Plant Manager - Hope Creek Attachment BJT C

Distribution LER File 3.7 1

4711%

11

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSIO APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0613012007 (6-2004)

, the NRC may digits/characters for each block) not conduct or sponsor, and a person Is not required to respond to, the

3. PAGE Hope Creek Generating Station 05000354 1 OF 3
4. TITLE Reactor Coolant System Leak from Check Valve Position Indicator
5. EVENT DATE
6. LER NUMBER
7. REPORT DATE
8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER

.RE MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 06 07 2005 2005 - 003 -

00 08 08 2005

9. OPERATING MODE
11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Checkll thatapply) o 20.2201(b) 0 20.2203(aX3)(i) 0 50.73(aX2)(i)(C) 0 50.73(a)(2)(vii) 1 0 20.2201(d) 0 20.2203(aX3)(3i) 0 50.73(aX2)(iiXA) 0 50.73(a)(2)(vIIIXA) ol 20.2203(aX1) 0 20.2203(a)(4) 0 50.73(aX2)(iiXB) 0 50.73(a)(2)(viii)(B) 0o 20.2203(aX2)(1) 0 50.36(c)(1XI)(A) 0 50.73(aX2)(1ij) 0 50.73(aX2)(Ix)(A)
10. POWER LEVEL 0 20.2203(a)(2)(1i) 0 50.36(c)(1Xii)(A) 0 50.73(aX2XIvXA) 0 50.73(aX2)(x) o 20.2203(a)(2Xii) 0 50.36(c)(2) 0 50.73(a)(2XvxA) 0 73.71(aX4) o 20.2203(a)(2)(iv) 0 50.46(a)(3#iI) 0 50.73(a)(2XvXB) 0 73.71(a)(5) 100 0 20.2203(a)(2)(v)

E0 50.73(a)(2XIXA)

D 50.73(aX2)(vXC) 0 OTHER 0 20.2203(aX2)(vi) 0 50.73(aX2)(i)(B) 0 50.73(aX2)(v)(D)

Specify In Abstract below nr In NRC. Fnrmq.A

12. LICENSEE CONTACT FOR THIS LER FACILlTY NAME TELEPHONE NUMBER (Inducde Area Code)

Brian Thomas, Licensing Engineer 856-339-2022

13. COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE

.._____FACTURER

,j TO EPIX FACTURER TO EPIX X

BO V

A585_

N

14. SUPPLEMENTAL REPORT EXPECTED
15. EXPECTED MONTH DAY YEAR SUBMISSION 0E YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE 09 30 2005 ABSTRACT (Limit to 1400 spaces, Le., approximately 15 single-spaced typewritten lines)

On June 7, 2005, at 1413 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.376465e-4 months <br />, the reactor recirculation pumps were taken to minimum speed and the reactor mode switch was locked in shutdown to scram the reactor due to increasing drywell pressure and drywell floor drain leakage. An Unusual Event (UE) was declared at 1437 hours0.0166 days <br />0.399 hours <br />0.00238 weeks <br />5.467785e-4 months <br /> due to drywell floor drain leakage exceeding 10 gpm.

The operating crew took action to bring the plant to a cold shutdown condition and a drywell entry was performed. At 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> on June 8, 2005, the source of the leak was determined to be from the F050A residual heat removal (RHR) check valve position indicator. On June 8, 2005, at 0336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br />, the B RHR loop was placed in shutdown cooling and Operational Condition 4 was entered at 0456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br />. The UE was terminated at 0515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br /> on June 8, 2005.

The cause of the reactor coolant system leak was the result of an approximately 285 degree circumferential crack in the position indicating tube for the F050A RHR check valve. The cause of the position indicating tube failure is still under investigation and will be reported in a supplement to this LER. Corrective actions taken consist of modifying the F050A and F050B RHR check valves to remove the position indicating tube and ultrasonic testing of the six other check valves with the same position indicating tubes. Additional corrective actions will be determined upon completion of the cause investigation. The supplemental LER will be submitted by September 30, 2005.

This event is being reported in accordance with I OCFR50.73(a)(2)(i)(A), "the completion of any nuclear plant shutdown required by the plant's Technical Specifications," 1 OCFR50.73(a)(2)(iv)(A), 'any event or condition that resulted in manual or automatic actuation of...reactor protection system (RPS) including: reactor scram or reactor trip," and 1 OCFR50.73(a)(2)(ii)(A), "any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.!

NRC FORM 366 (6-2004)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)

LICENSEE EVENT REPORT (LER)

FACILITY NAME (1)

DOCKET (2)

LER NUMBER (6)

PAGE (3)

_YEAR SEQUENTIAL I REVISION I A NUMBER NUMBER Hope Creek Generating Station 05000354 2 OF 3 1_

12005 003 00 TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor (BWRI4)

Reactor Coolant System {AB}*

Residual Heat Removal {BOI*

  • Energy Industry Identification System {EIIS) codes and componentfunction identifier codes appear as {SS/CCC)

IDENTIFICATION OF OCCURRENCE Event Date: June 7, 2005 Discovery Date: June 7, 2005 CONDITIONS PRIOR TO OCCURRENCE Hope Creek was in operational condition I at 100% power prior to the event. The B residual heat removal (RHR) pump was out of service for maintenance at the start of the event. There was no other equipment out of service that impacted this event.

DESCRIPTION OF OCCURRENCE On June 7, 2005, at 1351 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.140555e-4 months <br />, the operating crew entered the abnormal operating procedure for drywell leakage. At 1411 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.368855e-4 months <br /> with drywell pressure at 0.18 psig, the drywell leak detection system alarmed. At 1412 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.37266e-4 months <br /> with drywell pressure at 0.27 psig and continuing to rise, reactor power was reduced to 80% at the direction of the control room supervisor. At 1413 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.376465e-4 months <br /> with drywell pressure at 0.40 psig, the reactor recirculation pumps were reduced to minimum speed and the reactor mode switch was locked in shutdown to scram the reactor. Technical Specification (TS) 3.4.3.2 was entered for unidentified leakage greater than 5 gpm. TS 3.4.3.2 requires the reduction of leakage to within limits in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or to be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

At 1421 hours0.0164 days <br />0.395 hours <br />0.00235 weeks <br />5.406905e-4 months <br /> on June 7, 2005 with drywell pressure at 0.47 psig, the operating crew commenced the lowering of reactor pressure to 600 psig. At 1426 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.42593e-4 months <br />, drywell pressure was at 0.46 psig and drywell leakage was determined to be greater than 10 gpm. At 1427 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.429735e-4 months <br /> with reactor vessel pressure control being provided by the electro-hydraulic control (EHC) system, reactor water level 8 was reached. At 1429 hours0.0165 days <br />0.397 hours <br />0.00236 weeks <br />5.437345e-4 months <br />, drywell floor drain leakage was determined to be approximately 10.5 gpm with drywell pressure at 0.43 psig. At 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br />, reactor vessel water level 3 scram level was reached due to the reactor vessel pressure reduction. Reactor vessel pressure was at 600 psig. The operating crew further reduced pressure to 550 psig. During the pressure reduction, reactor vessel water level reached level 8. At 1434 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.45637e-4 months <br /> with reactor vessel pressure at 550 psig, drywell floor drain leakage was approximately 11.3 gpm and drywell pressure was 0.40 psig. At 1437 hours0.0166 days <br />0.399 hours <br />0.00238 weeks <br />5.467785e-4 months <br />, an Unusual Event (UE) was declared due to drywell floor drain leakage being greater the 10 gpm. At 1438 hours0.0166 days <br />0.399 hours <br />0.00238 weeks <br />5.47159e-4 months <br />, the operating crew commenced the cool down of the reactor in accordance with procedures. Reactor vessel level was being maintained at +35". At 1442 hours0.0167 days <br />0.401 hours <br />0.00238 weeks <br />5.48681e-4 months <br />, drywell floor drain leakage was approximately 14.6 gpm with drywell pressure at 0.37 psig. At 1458 hours0.0169 days <br />0.405 hours <br />0.00241 weeks <br />5.54769e-4 months <br />, drywell floor drain leakage was approximately 11.2 gpm with reactor pressure at 500 psig. At 1522 hours0.0176 days <br />0.423 hours <br />0.00252 weeks <br />5.79121e-4 months <br />, drywell floor drain leakage was steady at 9.9 gpm. At 1618 hours0.0187 days <br />0.449 hours <br />0.00268 weeks <br />6.15649e-4 months <br />, reactor pressure was reduced to 300 psig. At 1651 hours0.0191 days <br />0.459 hours <br />0.00273 weeks <br />6.282055e-4 months <br />, the operating crew reset the scram signal. The position indication for the F050A residual heat removal (RHR) {BONV) check valve was lost at 1802 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.85661e-4 months <br />.

At 1823 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.936515e-4 months <br /> the B RHR pump was restored from maintenance and retested satisfactorily. At 2344 hours0.0271 days <br />0.651 hours <br />0.00388 weeks <br />8.91892e-4 months <br /> with reactor pressure less than 82 psig, the A RHR train was declared inoperable for shutdown cooling due to questions with the failure of the valve position indication. Preparations were made for drywell entry and placing B RHR in shutdown coolina.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)

LICENSEE EVENT REPORT (LER)

FACILITY NAME (1)

DOCKET (2)

LER NUMBER (6)

PAGE (3)

SEQUENTIAL REVISION Hope Creek Generating Station 05000354 3 OF 3

__2005 003 00 DESCRIPTION OF OCCURRENCE (cont'd)

At 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> on June 8, 2005, a report from the drywell stated that the leakage was coming from the F050A RHR check valve position indicator. At 0335 hours0.00388 days <br />0.0931 hours <br />5.539021e-4 weeks <br />1.274675e-4 months <br /> B RHR was placed in service for shutdown cooling. The plant entered Operational Condition 4 (cold shutdown) at 0456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br />. The UE was terminated at 0515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br /> on June 8, 2005.

This event is being reported in accordance with 1 OCFR50.73(a)(2)(i)(A), "the completion of any nuclear plant shutdown required by the plant's Technical Specifications," 1 OCFR50.73(a)(2)(iv)(A), "any event or condition that resulted in manual or automatic actuation of...reactor protection system (RPS) including: reactor scram or reactor trip," and 1 OCFR50.73(a)(2)(ii)(A), "any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.t CAUSE OF OCCURRENCE The cause of the reactor coolant system leak was the result of an approximately 285 degree circumferential crack in the position indicating tube for the F050A RHR check valve. The cause of the position indicating tube failure is still under investigation and will be reported in a supplement to this LER. The supplemental LER will be submitted by September 30, 2005.

PREVIOUS OCCURRENCES

A review for prior similar occurrences will be performed upon completion of the cause investigation.

SAFETY CONSEQUENCES AND IMPLICATIONS

On June 7, 2005, drywell floor drain leakage increased to greater than 10 gpm. As a result an Unusual Event (UE) was declared. The operating crew took appropriate actions to bring the plant to a controlled cold shutdown condition. Reactor water level was maintained above level 2 (initiation signal for emergency core cooling injection) and reactor pressure control was maintained throughout the event. Even if a complete severance of the F050A RHR check valve position indicating tube were to occur, this failure would not result in an uncontrolled reactor depressurization. Based on the above, there was no impact to the health and safety of the public.

A review of this event determined that a Safety System Functional Failure (SSFF) has not occurred as defined in Nuclear Energy Institute (NEI) 99-02. Hope Creek was brought to a cold shutdown condition following the identification of the reactor coolant system leak from the position indicating tube of the F050A RHR check valve.

CORRECTIVE ACTION

I.

The F050A and F050B RHR check valves were modified to remove to remove the position indicator tubes. These modifications were completed prior to the plant startup.

2. Six other check valves contain the same position indicator tube as the F050A and FOSOB RHR check valves. The position indicating tubes for these six check valves underwent ultrasonic testing with no indications of wear as experienced on the F050A RHR check valve. These inspections were completed prior to the plant startup.

Additional corrective actions will be determined upon completion of the event investigation. This report will be supplemented by September 30, 2005.

COMMITMENTS

This LER contains no commitments.

Retrieved from "https://kanterella.com/w/index.php?title=05000354/LER-2005-003,_Regarding_Reactor_Coolant_System_Leak_from_Check_Valve_Position_Indicator&oldid=5378258"