05000327/LER-2009-005, Regarding Manual Reactor Trip Following a Loss of Flow Through Loop 1 Feedwater Regulating Valve

Regarding Manual Reactor Trip Following a Loss of Flow Through Loop 1 Feedwater Regulating Valve

ML091871021
Person / Time
Site:	Sequoyah  (DPR-077)
Issue date:	07/06/2009
From:	Clearly T Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 09-005-00
Download: ML091871021 (7)

LER-2009-005, Regarding Manual Reactor Trip Following a Loss of Flow Through Loop 1 Feedwater Regulating Valve

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2) 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x)
3272009005R00 - NRC Website

text

Tennessee Valley Authority Post Office Box 2000 Soddy Daisy, Tennessee 37384-2000 Timothy P. Cleary Site Vice President Sequoyah Nuclear Plant July 6, 2009 U.S. Nuclear Regulatory Commission 10 CFR 50.73 ATTN:

Document Control Desk Washington, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY - SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 1

- DOCKET NO.

50-327 - FACILITY OPERATING LICENSE DPR LICENSEE EVENT REPORT (LER) 50-327/2009-005-00 The enclosed LER provides details concerning a manual reactor trip and engineered safety feature actuation of auxiliary feedwater.

The manual trip was initiated because of a low and decreasing steam generator level when a feedwater regulating valve closed as a result of a ruptured actuator diaphragm.

This report is being submitted in accordance with 10 CFR 50.73(a)(2)(iv)(A), as an event that resulted in an automatic actuation of the reactor protection system.

Sincerely, Timothy P. Cleary Enclosure cc:

See page 2 printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 July 6, 2009 Enclosure cc (Enclosure):

INPO Records Center Institute of Nuclear Power Operations 700 Galleria Parkway, SE, Suite 100 Atlanta, Georgia 30339-5957 Mr. Siva P. Lingam, Project Manager U.S. Nuclear Regulatory Commission Mail Stop 08G-9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379

NRC FORM 366 (9-2007)

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

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

APPROVED BY OMB NO.

3150-0104 EXPIRES 08/31/2010 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 (T-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, (3150-0104),

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 required to respond to, the information collection.

1. FACILITY NAME Sequoyah Nuclear Plant (SQN) Unit 1

2. DOCKET NUMBER 05000327

3. PAGE 1 OF 5

4. TITLE:

Manual Reactor Trip Following a Loss of Flow Through Loop 1 Feedwater Regulating Valve

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.

MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05 06 2009 2009 005 00 07 06 2009 FACILITY NAME DOCKET NUMBER

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §:

10. POWER LEVEL 100

20.2201 (b)

20.2201 (d)

20.2203(a)(1)

20.2203(a)(2)(l)

20.2203(a)(2)(ii)

20.2203(a)(2)(iii)

20.2203(a)(2)(iv)

20.2203(a)(2)(v)

20.2203(a)(2)(vi)

20.2203(a)(3)(i)

20.2203(a)(3)(ii)

20.2203(a)(4)

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

50.36(c)(1)(ii)(A)

50.36(c)(2)

50.46(a)(3)(ii)

50.73(a)(2)(i)(A) 50.73(a)(2)(i)(B)

50.73(a)(2)(i)(C) 50.73(a)(2)(ii)(A) 50.73(a)(2)(ii)(B) 50.73(a)(2)(Jii) 50.73(a)(2)(iv)(A) 50.73(a)(2)(v)(A) 50.73(a)(2)(v)(B) 50.73(a)(2)(v)(C) 50.73(a)(2)(v)(D)

(Check all that apply)

50.73(a)(2)(vii)

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

50.73(a)(2)(vJii)(B)

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

50.73(a)(2)(x)

73.71 (a)(4)

73.71 (a)(5)

OTHER Specify in Abstract below or in (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A)

VII.

ADDITIONAL INFORMATION

A.

Failed Components:

The failed component was a diaphragm in a Fisher Controls International Inc. Size 80 Type 667 actuator in the Loop 1 main feedwater regulating valve.

The diaphragm failure was determined to be improper clamping force of the diaphragm plates that support and attach the diaphragm to the actuator stem.

This insufficient clamping force was a result of insufficient torque applied to the fastening cap screw.

The insufficient clamping force allowed stress concentrations at the diaphragm hole to be approximately three times nominal values, which caused an initial tear in the diaphragm composite material and led to the instantaneous failure of the diaphragm.

All equipment responded as required except the failed main feedwater regulating valve had a "not closed" indication in the main control room following a feedwater isolation signal.

B.

Previous LERs on Similar Events:

A review of previous reportable events for the past 10 years indicated that there have been reactor trips regarding main feedwater regulating valves; however, there are no previous events that were a result of a diaphragm failure in a feedwater regulating valve.

C.

Additional Information

None.

D.

Safety System Functional Failure:

This event did not result in a safety system functional failure in accordance with 10CFR50.73(a)(2)(v).

E.

Unplanned Scram with Complications:

This condition did not result in an unplanned scram with complications.

VIII.

COMMITMENTS

None.