05000260/LER-2009-007, Regarding Manual Scram During Removal of a Reactor Feedwater Pump from Service

Regarding Manual Scram During Removal of a Reactor Feedwater Pump from Service

ML093280854
Person / Time
Site:	Browns Ferry
Issue date:	11/20/2009
From:	West R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 09-007-00
Download: ML093280854 (6)

LER-2009-007, Regarding Manual Scram During Removal of a Reactor Feedwater Pump from Service

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(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
2602009007R00 - NRC Website

text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 November 20, 2009 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Browns Ferry Nuclear Plant, Unit 2 Facility Operating License No. DPR-52 NRC Docket No. 50-260

Subject:

Licensee Event Report 50-260/2009-007, "Manual Scram During Removal of a Reactor Feedwater Pump from Service" The submittal provides Licensee Event Report (LER) 260/2009-007. The LER provides the details of a manual reactor scram that was inserted during work activities associated with removing a reactor feedwater pump from service. TVA is reporting this occurrence in accordance with 10 CFR 50.73(a)(2)(iv)(A).

There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact F. R. Godwin, Site Licensing and Industry Affairs Manager, at (256) 729-2636.

espectfully, R. G. West Vice President Enclosure cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 08/31/2010 (9-2007)

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3.

AGE Browns Ferry Nuclear Plant Unit 2 05000260 1 of 5

4. TITLE: Manual Scram During Removal of a Reactor Feedwater Pump from Service

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED SEAUENTIAL REV FACILITY NAME DOCKET NUMBER DA E R Y A U BER NO.

N n

/

FACILITY NAME DOCKET NUMBER 09 29 2009 2009 -

007 00 11 20 2009 None N/A

9. OPERATING MODE

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

[1 20.2201(b)

El 20.2203(a)(3)(i)

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

El 50.73(a)(2)(vii)

Dl 20.2201(d)

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

[E 50.73(a)(2)(ii)(A)

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

El 20.2203(a)(1)

El 20.2203(a)(4)

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

[I 50.73(a)(2)(viii)(B)

[I 20.2203(a)(2)(i)

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

El 50.73(a)(2)(iii)

El 50.73(a)(2)(ix)(A)

10. POWER LEVEL El 20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)

[I 50.73(a)(2)(x)

El 20.2203(a)(2)(iii)

El 50.36(c)(2)

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

[1 73.71(a)(4)

El 20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

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

El 73.71 (a)(5) 100 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

D. Other Systems or Secondary Functions Affected

None.

E. Method of Discovery

Operators observed low pump suction pressure alarms, pump trips, and decreasing reactor vessel water level.

F. Operator Actions

Operator actions in response to the scram were in accordance with applicable operating procedures.

G. Safety System Responses Safety systems operated as designed. The RIC system, although not a safety system, did not operate as designed.

II1. CAUSE OF THE EVENT A.

Immediate Cause The immediate cause of the event was the successive trips of reactor feedwater pump 2A, condensate booster pump 2A, and reactor feedwater pump 2C.

B. Root Cause The cause investigation determined that the condensate/feedwater system operating instructions had been previously revised to allow operation at 100 percent power in a reduced pump configuration. The basis for the procedure change was the misapplication of a steady-state hydraulic design calculation.

C. Contributing Factors The risk review process did not adequately identify the aggregate risk (reduction in operating margin) of removing a reactor feedpump from service with other condensate system pumps out-of-service.

IV. ANALYSIS OF THE EVENT

There are three pairs of electric condensate and condensate booster pumps, which provide flow and pressure head to the three steam driven reactor feedwater pumps on each BFN unit. During normal operation, all nine pumps are in service. On Units 1 and 2 the original pumps have been replaced with larger capacity pumps to support extended power uprate (EPU) operation.

Prior to this event, condensate pumps 2A and 20, condensate booster pumps 2A and 2B, and all three reactor feedwater pumps were operating. Condensate pump 2B and condensate booster pump 20 had been previously removed from service for scheduled maintenance. At 2321 hours0.0269 days <br />0.645 hours <br />0.00384 weeks <br />8.831405e-4 months <br />, operators were in process of reducing speed and flow on feedwater pump 2B to take it off line for scheduled maintenance. As feedwater pump 2B speed was being lowered, reactor feedwater pumps 2A and 2C automatically increased speed and flow to maintain rated feedwater flow to the reactor.

When the speed of feedwater pump 2B was lowered to where pump flow was below the minimum flow setpoint, feedwater pump 2B minimum flow valve automatically opened, which opened a flow path back to the condenser. This resulted in increasing total condensate flow, which reduced the head pressure being supplied to the operating reactor feedwater pumps and condensate booster pumps.

This was directly evidenced by the receipt of low pump suction pressure alarms. The low header pressure also initiated the low pump suction pressure trip logic that provides trips to the feedwater pumps and condensate booster pumps after staggered time delays.

Operators attempted to recover suction head by manually increasing the speed of feedwater pump 2B, which would cause the 2B minimum flow valve to reclose, and by decreasing reactor power by reducing recirculation pump speed. However, the low pump suction pressure trip logic timed out and feedwater pump 2A and condensate booster pump 2A tripped. Feedwater pump 2C attempted to maintain reactor water level and tripped on overspeed shortly after booster pump 2A tripped.

With decreasing reactor vessel water level, operators manually scrammed the reactor at approximately +15 inches above instrument zero on the normal operating range water level instrumentation. The automatic scram setpoint on reactor water level is at +2 inches. As reactor water level continued to decrease because of the feedwater pump trips and due to the prompt water level decrease from the scram, the HPCI and RCIC systems auto-initiated on Reactor Vessel Low Low, Level 2 (-45 inches). HPCI and feedwater pump 2B operation restored reactor water level to normal within approximately 45 seconds after the scram. RCIC started, but failed to achieve sufficient speed and discharge pressure to inject.

The post-event evaluation determined that the reactor control systems and feedwater/condensate systems responded as would be expected for the circumstances.

The cause investigation determined that the condensate/feedwater system operating instructions had been previously revised to allow plant operation at 100 percent power in a reduced pump configuration status (with a condensate, a condensate booster pump, and a feedwater pump out-of-service). The basis for the procedure change was a misapplication of a hydraulic flow calculation performed for EPU. The EPU calculation was a steady-state analysis that showed 100 percent power could be achieved with the reduced pump configuration; however, the calculation did not address the transient effects of establishing the reduced pump configuration. In this case, the revised operating instructions did not account for the additional condensate system flow expected when a feedwater pump minimum flow valve opens. The cause investigation also determined that risk review process for conducting online feedwater/condensate system maintenance did not adequately identify the risk (reduction in operating margin) associated with removing a feedwater pump from service for maintenance with other condensate system pumps out-of-service.

V. ASSESSMENT OF SAFETY CONSEQUENCES

The trip of the feedwater pumps and condensate booster pump resulted in an unplanned scram and the initiation of high pressure injection systems. The HPCI system responded to the transient and reactor water level was restored to a safe condition by the operation of HPCI and feedwater pump 2B, which remained in service during the entire event. As noted previously, RCIC failed to inject. The main condenser remained in service throughout and served as the heat sink for the reactor. The event is categorized as a partial loss-of-feedwater event. The plant is analyzed for a complete loss-of-feedwater event.

VI. CORRECTIVE ACTIONS

A.

Immediate Corrective Actions

The operating instructions were revised to require that feedwater flow be below that needed for 85 percent power prior to removing a feedwater pump from service if a condensate or condensate booster is not in service. With the lower feedwater flow, adequate pump suction head

would be maintained during the removal of a reactor feedwater pump from service. Additionally, a new tool for assessing the risk of online activities was implemented.

B. Corrective Actions to Prevent Recurrence - The corrective actions are being managed by the Browns Ferry Nuclear Plant corrective action program.

The risk review process for performing online maintenance on the feedwater/condensate system with a reduced complement of pumps in service will also be reviewed.

VII. ADDITIONAL INFORMATION

A. Failed Components None.

B.

Previous LERs on Similar Events None identified.

C. Additional Information

The primary corrective action documents for this event are Problem Evaluation Report (PER) 203538 (scram event) and PER 171722 (evaluate risk process).

D. Safety System Functional Failure Consideration:

This event did not result in a safety system functional failure according to NEI 99-02.

E. Scram With Complications Consideration:

This event was not a complicated scram according to NEI 99-01.

VIII. COMMITMENTS

None.