05000282/LER-2008-003, Loss of Auxiliary Feedwater Safety Function and Condition Prohibited by Technical Specifications Due to Mispositioned Isolation Valve

Loss of Auxiliary Feedwater Safety Function and Condition Prohibited by Technical Specifications Due to Mispositioned Isolation Valve

ML082730902
Person / Time
Site:	Prairie Island
Issue date:	09/29/2008
From:	Wadley M Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-08-077, LER 1-08-03 LER 08-003-00
Download: ML082730902 (6)

LER-2008-003, Loss of Auxiliary Feedwater Safety Function and Condition Prohibited by Technical Specifications Due to Mispositioned Isolation Valve

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 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
2822008003R00 - NRC Website

text

September 29, 2008 L-PI-08-077 10 CFR 50.73 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Unit 1 Docket 50-282 License No. DPR-42 LER 1-08-03, Loss of Auxiliarv Feedwater Safety Function and Condition Prohibited by Technical Specifications Due to Mispositioned Isolation Valve Northern States Power Company, a Minnesota corporation (NSPM) herewith encloses Licensee Event Report (LER) 1-08-03. The LER describes a condition where one train of the auxiliary feedwater (AFW) system was inadvertently rendered inoperable by a mispositioned isolation valve on a pump discharge pressure switch. The condition resulted in one train of AFW being inoperable longer than allowed by technical specifications and in a loss of all AFW function when the opposite train of AFW was concurrently inoperable due to surveillance testing. Please contact us if you require additional information related to this event.

Summary of Commitments This letter contains no new commitments and no changes to existing commitments.

Michael D. Wadley Site Vice President i

Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:

Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC Department of Commerce, State of Minnesota 171 7 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1 121

ENCLOSURE LICENSEE EVENT REPORT 1-08-03 4 Pages Follow

LICENSEE EVENT REPORT (LER)

I (See reverse for required number of digitslcharacters for each block)

I.

FACILITY NAME Prairie Island Nuclear Generating Plant Unit 1 I

4. TITLE Loss of AFW Safety Function and Condition Prohibited by Technical Specifications Due to Mispositioned Isolation Valve Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-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 (l-5 F52), 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, (3150-0066), 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 required to respond to, the information collection.

1

5. EVENT DATE I

6. LER NUMBER 1

7. REPORT DATE I

8. OTHER FACILITIES INVOLVED

2. DOCKET NUMBER 05000282

3. PAGE 1 of4 20.2203(a)(2)(i) 50.36(c)(l)(i)(A)

[7 50.73(a)(2)(iii)

[7 50.73(a)(2)(ix)(A)

10. POWER LEVEL H

20.2203(a)(2)(ii) 50,36(c)(l)(ii)(A) 50,73(a)(2)(iv)(A) 50,73(a)(2)(x)

MONTH 7

I 20.2203(a)(2)(iii)

[7 50.36(~)(2) 50,73(a)(2)(v)(A) 73.71 (a)(4) 20.2203(a)(2)(iv)

[7 50,46(a)(3)(ii) a 50.73(a)(2)(v)(B)

[7 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)

[7 OTHER DAY 31

9. OPERATING MODE I

20.2203(a)(2)(vi) 50,73(8)(2)(i)(B) 50.73(a)(2)(v)(D)

Spec~fy in Abstract below or In NRC F w

266A

12. LICENSEE CONTACT FOR THlS LER

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check all that apply)

YEAR 2008 NAME Jeff Kivi 1 I I

I I

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

On July 31, 2008, Prairie Island Nuclear Generating Plant (PINGP) Unit 1 was operating at 100 percent power. At 0817 CDT Unit 1 reactor tripped (see LER 1-08-02 for details). Subsequent to the trip, the Unit 1 turbine-driven auxiliary feedwater pump (1 1 TDAFWP) auto started as designed, but tripped 42 seconds later on low discharge pressure. Local and remote indications did not indicate any abnormal operating conditions.

A prompt investigation into 11 TDAFWP trip revealed the discharge pressure switch (PS-17700) was isolated.

A time delay in the pump protective circuitry is designed to trip the pump on a continuing low discharge pressure condition. Monthly surveillances to test the operability of the 11 TDAFWP do not test the low discharge pressure trip function of the pump as the pump is in manual, which bypasses the low discharge pressure trip. The 12 motor-driven auxiliary feedwater pump (1 2 MDAFWP) started as designed and ran as needed throughout the event.

TELEPHONE NUMBER (Include Area Code) 651.388.1 121 I

Valve lineups were verified, the 11 TDAFWP was tested and restored to operable status. The subject valve was lock wired into the open position. Action will be taken to identify any similar components that can affect operability of safety systems and to revise procedures for locking and blocking.

YEAR REPORTABLE TO EPlX I

NRC FORM 366 (9-2007)

I 1 SEQUENTIAL NUMBER I

14. SUPPLEMENTAL REPORT EXPECTED 0

YES (If yes, complete 15. EXPECTED SUBMISSION DATE).

O NO 2008 - 003 - 00 MANU-FA CTURER SYSTEM

CAUSE

REV NO

CAUSE

COMPONENT I

COMPONENT SYSTEM 09 YEAR

15. EXPECTED SUBMISSION DATE MONTH MANU-FACTURER 29 REPORTABLE TO EPlX MONTH DAY DAY 2008 YEAR FACILITY NAME DOCKET NUMBER FACILITY NAME DOCKET NUMBER

EVENT DESCRIPTION LICENSEE EVENT REPORT (LER)

U.S. NUCLEAR REGULATORY COMMISSION (9-2007)

CONTINUATION SHEET On July 31, 2008, Prairie Island Nuclear Generating Plant (PINGP) Unit 1 was operating at 100 percent power. At 081 7 CDT Unit 1 reactor' tripped (see LER 1-08-02 for details). Subsequent to the trip, the Unit 1 turbine-driven auxiliary feedwate? pump3 (1 1 TDAFWP) auto started as designed, but tripped 42 seconds later on low discharge pressure. Local and remote indications did not indicate any abnormal operating conditions. A prompt investigation into 11 TDAFWP trip revealed the discharge pressure switch4 (PS-17700) was isolated. A time delay in the pump protective circuitry is designed to trip the pump on a continuing low discharge pressure condition. The 12 motor-driven auxiliary feedwater pump (12 MDAFWP) started as designed and ran as needed throughout the event.

I. FACILITY NAME Prairie Island Nuclear Generating Plant Unit 1 At the end of the Unit 1 outage (1 R25) in February 2008, surveillance SP 1234A was performed on 11 TDAFWP to calibrate the discharge pressure switch on the pump. The final position of the isolation valve5 to the switch was open and independently verified. A valve lineup verified the position on March 11, 2008. Surveillance SP 1301 was performed to test all trip and safety related functions of 11 TDAFWP and SP 1376 was performed to verify operation of system valves. SP 1103 was performed to verify post-outage AFWP flow. During each of the four subsequent months, SP 1102 was also performed to test manual pump operation (with the pump in manual, the low discharge pressure trip is bypassed, so SP 1102 would not be expected to identify the as-found condition).

None of the monthly procedures manipulated the PS-17700 isolation valve, but all required the operation of isolation valves in close proximity and of the same type as the isolation valve to the discharge pressure switch.

EVENT ANALYSIS

2. DOCKET NUMBER 05000282 PINGP Technical Specification (TS) 3.7.5 requires both trains of the auxiliary feedwater system (AFW) to be operable in Modes 1, 2, 3, and Mode 4 when the steam generator is relied upon for heat removal. TS 3.7.5 Condition B requires that an inoperable train is to be restored in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If Condition B is not met, Condition C requires the unit be shut down to Mode 3 in six hours and to Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Sometime between March 11, 2008, and July 31, 2008, the isolation valve for PS-17700 was closed rendering the 11 TDAFWP inoperable. The available evidence indicates the valve was likely mispositioned during one of the periodic surveillances before the reactor trip, which would indicate Unit 1 remained in Mode 1 with one train of AFW inoperable for greater than 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />.

Thus, the as-found condition of the 11 TDAFWP is reportable as a condition prohibited by TS, per 10 CFR 50.73(a)(2)(i)(B).

1 Ells Component Identifier: RCT Ells System Code: BA 3

4 Ells Component Identifier: P Ells Component Identifier: PS 5 Ells Component Identifier: ISV

6. LER NUMBER YEAR SEQUENTIAL REV NUMBER NO 2008 -

003

- 00

3. PAGE 2 o f 4

Additionally, the available evidence indicates it is likely the period the 11 TDAFWP was inoperable (due to the isolation of PS-17700) overlapped one of the occasions that the 12 MDAFWP was inoperable for monthly surveillance testing. The AFW System automatically supplies feedwater to the steam generators to remove decay heat from the Reactor Coolant System upon the loss of normal feedwater supply. The periods of overlapping inoperability of both trains of AFW are reportable as a condition that could have prevented the fulfillment of a safety function needed to remove residual heat per 10 CFR 50.73(a)(2)(v)(B). LICENSEE EVENT REPORT (LER)

U.S. NUCLEAR REGULATORY COMMISSION (9-2007)

CONTINUATION SHEET

SAFETY SIGNIFICANCE

The exact amount of time the 11 TDAFWP was inoperable due to the isolated PS-17700 is not known, but can be bounded from the last time position of the isolation valve was independently verified. For most of this time, the redundant 12 MDAFWP was available to meet the safety function.

For the periods that both trains of AFW were inoperable, operators could have restored one or both trains had an actual system demand occurred. Either the 12 MDAFWP could have been restored from its surveillance testing configuration or the 11 TDAFWP could have been restored per the alarm response procedures. Thus, this event did not affect the health and safety of the public and the safety significance of this event is considered minimal.

3. PAGE 3 of 4

CAUSE

I

6. LER NUMBER YEAR SEQUENTIAL REV NUMBER NO 2008 -

003

- 00 I. FACILITY NAME Prairie Island Nuclear Generating Plant Unit 1 An evaluation of the cause of the isolation of the 11 TDAFWP discharge pressure switch revealed the root cause was a failure of the site to adequately control components that affect safety related equipment.

The mispositioned valve was not labeled, bypassing barriers to make identification more likely.

The procedure for safeguards hold cards and component blocking or locking contains a definition of what components should not be controlled, but does not contain a definition of which components should be controlled per the safeguards hold card program.

The mispositioned valve was not locked in the required position, making mispositioning more likely. The procedure does not define which valves shall have locks, blocks or lock wires installed.

2. DOCKET NUMBER 05000282 The root cause evaluation identified a contributing cause of an unrecognized operator burden that created an error likely situation. To correct a condition during specific conditions that over-range the turbine driven pump suction pressure gage6, several procedures were changed that kept the gage isolated, except when needed. This action required additional valve manipulations in the vicinity of the PS-17700 manifold isolation valve.

As part of the initial investigation into this event, the site determined that tampering was not evident with this incident.

6 Ells Component Identifier: PI

CORRECTIVE ACTION LICENSEE EVENT REPORT (LER)

U.S. NUCLEAR REGULATORY COMMISSION (9-2007)

CONTINUATION SHEET Surveillance procedures were completed to verify AFW system operability. The positions of other valves with similar functions in AFW system have been checked. The suction and discharge pressure switch manifold isolation valves for all four AFW pumps were lock-wired in the open position. A sample of valves on other systems was walked down to verify the positions of valves with similar functions - no discrepancies were noted.

Additional corrective actions include:

3. PAGE 4 o f 4

1. FACILITY NAME Prairie Island Nuclear Generating Plant Unit 1

1. Conduct a review to find all other components that could affect operability of safety related systems. Components to be included in equipment database and drawings, locking devices be installed and components to be labeled in the field.

2. Revise the procedure for component locking and blocking.

PREVIOUS SIMILAR EVENTS

2. DOCKET NUMBER 05000282 LER 1-06-02 and LER 1-08-01 report instances of inoperability of a train of AFW. The causes of these previous events were substantially different from the event of this LER, thus, the corrective actions from the previous events would not be expected to prevent the event of this LER.

6. LER NUMBER YEAR SEQUENTIAL REV NUMBER NO 2008 -

003

- 00