05000389/LER-2009-004, Unplanned Manual Reactor Trip During Reactor Startup

Unplanned Manual Reactor Trip During Reactor Startup

ML093280855
Person / Time
Site:	Saint Lucie
Issue date:	11/19/2009
From:	Richard Anderson Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2009-248 LER 09-004-00
Download: ML093280855 (4)

LER-2009-004, Unplanned Manual Reactor Trip During Reactor Startup

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(iv), System Actuation
3892009004R00 - NRC Website

text

0 FPL Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 November 19, 2009 L-2009-248 10 CFR 50.73 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Re:

St. Lucie Unit 2 Docket No. 50-389 Reportable Event: 2009-004 Date of Event: September 21, 2009 Unit 2 Unplanned Manual Reactor Trip During Reactor Startup The attached Licensee Event Report 2009-004 is being submitted pursuant to the requirements of 10 CFR 50.73 to provide notification of the subject event.

Respectfully, Richard L. Anderson Site Vice President St. Lucie Plant RLA/dlc Attachment IT76~

(~4IZA~

an FPL Group company

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/201(

9-2007)

Estimated burden per response to comply with this mandatory collection request: 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52),

LICENSEE EVENT REPORT (LER)

U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to infocollects@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not I required to respond to. the information collection.

13. PAGE St. Lucie Unit 2 05000389 1

OF 3

4. TITLE Unit 2 Unplanned Manual Reactor Trip During Reactor Startup

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITYNAME DOCKET NUMBER IFACILITY NAME DOCKET NUMBER 09 21 2009 2009 -

004 00 11 19 2009

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)

[I 20.2201(b)

ED 20.2203(a)(3)(i)

El 50.73(a)(2)(i)(C)

E) 50.73(a)(2)(vii) 2 20.2201(d)

[I 20.2203(a)(3)(ii)

El 50.73(a)(2)(ii)(A)

El 50.73(a)(2)(viii)(A)

El 20.2203(a)(1)

El 20.2203(a)(4)

El 50.73(a)(2)(ii)(B)

El 50.73(a)(2)(viii)(B)

_ 20.2203(a)(2)(i)

El 50.36(c)(1)(i)(A)

El 50.73(a)(2)(iii)

[3 50.73(a)(2)(ix)(A)

10. POWER LEVEL El 20.2203(a)(2)(ii)

[I 50.36(c)(1)(ii)(A)

Z 50.73(a)(2)(iv)(A)

El 50.73(a)(2)(x)

El 20.2203(a)(2)(iii)

[] 50.36(c)(2)

El 50.73(a)(2)(v)(A)

El 73.71 (a)(4) 0%

El 20.2203(a)(2)(iv)

[I 50.46(a)(3)(ii)

El 50.73(a)(2)(v)(B).

E 73.71(a)(5)

El 20.2203(a)(2)(v)

El 50.73(a)(2)(i)(A)

El 50.73(a)(2)(v)(C)

[E OTHER El 20.2203(a)(2)(vi)

El 50.73(a)(2)(i)(B)

El 50.73(a)(2)(v)(D)

Specify in Abstract below or in NRC Fnrm 366A

12. LICENSEE CONTACT FOR THIS LER NAME

]TELEPHONE NUMBER (include Area Code)

Donald L.

Cecchett - Licensing Engineer 7

772-467-7155CAS YTMCMOET MANUI-REPORTABLE CAS YTMMANU-REPORTABLE

CAUSE

COMPONENT FACTURER TO EPIX

CAUSE

COMPONENT FACTURER TO EPIX B

AB PCV C710 YES

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE)

[

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

On September 21,

2009, St. Lucie Unit 2 was performing a plant startup following a 2B2 Reactor Coolant Pump maintenance outage.

St. Lucie Unit 2 was entering Mode 2 at 1325 EDT when a Control Room Operator noted that primary safety valve (PSV)

V1202 had indications of seat leakage based on increasing tailpipe temperatures.

At 1333 on 9/21/09, the Control Room operators observed V1202 tailpipe temperature increase to approximately 224 'F coincident with a quench tank pressure increase from 4.0 psig to 4.4 psig.

Subsequently, the Unit Supervisor ordered a manual reactor trip in accordance with plant procedures.

Note that the reactor had not yet achieved criticality.

The PSV's seat leakage was a result of the valve not achieving thermal stability early enough to resist the increasing pressure loading.

Contributing causes

included gas management during the startup process, missing pressurizer nozzle insulation, and procedure deficiencies.

Corrective actions included configuration changes to the PSV discharge piping, and procedure revisions to address start up rates and PSV thermal stability.

NCR FORM 366 (9-2007)

Description of the Event On September 20, 2009, St. Lucie Unit 2 was performing a plant startup following a 2B2 reactor coolant pump (RCP)

[EIIS: AB] maintenance outage.

St. Lucie Unit 2 was entering Mode 2 when a Control Room Operator observed that the primary safety valve (PSV)

[EIIS: AB] V1202 had indications of seat leakage based on increasing tailpipe temperatures.

Quench tank [EIIS: AB] levels did not increase, indicating that the leakage was only non-condensable gas from the pressurizer steam space.

Control Room operators stopped the leakage by reducing reactor coolant system (RCS)

[EIIS: AB]

pressure by 200 psi and slowly increasing the pressure over the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to minimize thermal loading on the valve.

Full RCS pressure of 2250 psia was reached at approximately 0900 on 9/21/09 with all pressurizer safety valve temperatures stable; reactor start-up commenced at 1227 on 9/21/09.

At 1333 on 9/21/09, the safety valve V1202 tailpipe temperature abruptly increased to approximately 224 0F coincident with quench tank pressure increasing from 4.0 psig to 4.4 psig.

Control Room Operators stopped the leakage by lowering RCS pressure to 2050 psia and the Unit Supervisor ordered a manual reactor trip.

All safe shutdown equipment operated as designed with no adverse impact on the health and safety of the public.

The reactor did not yet achieved criticality.

Cause of the Event

An evaluation determined the causes for this event can be attributed to PSV seat leakage resulting from the valve not achieving thermal stability early enough to resist the increasing pressure.

Contributing causes included gas management during the startup process, missing PSV inlet flange insulation, and procedure deficiencies.

Analysis of the Event

This event is reportable under 10 CFR 50.73(a) (2) (iv)

(A),

as any event or condition that resulted in a manual or automatic reactor trip.

Analysis of Safety Significance The PSVs are installed to provide over pressure protection for the RCS.

The PSVs provide an active safety function which provides RCS overpressure protection during increasing pressure transients in accordance with ASME Boiler and Pressure Vessel Code Section III.

The PSVs also provide a passive safety function to maintain the RCS pressure boundary.

Actions taken by the Operators to manually trip the Unit precluded any safety concerns for the PSVs to seat during startup; safe shutdown was achieved without impacting the health and safety of the public.

Corrective Actions

The following corrective and supporting actions were entered into the site corrective action program.

1. Replaced PSV valve V1202 with an identical model valve.

2. Added insulation collars on all three PSV inlet pipe flanges.

3.

Upgrade PSV removal and installation maintenance procedure 0-MMP-01.09

4. Revise Operation procedure 2-GOP-502 to mandate slow RCS pressurization rate and annotate these changes to sustain actions for sustainability (PSV thermal stability issue)

Similar Events

Review of operating history for St. Lucie identified five events of seat leakage from the Pressurizer Code Safety Relief Valves.

The causes varied; this is a repeat event.

Failed Components Crosby Valve and Gage Company pressure relief valves; The valves are stainless steel forged body model DS-C-84217 size 3K6 (3"

inlet and 6" outlet) with a 3" 2500 # large tongue inlet flange, and a 6" 300# raised face exit flange.