05000364/LER-2016-001, Regarding Manual Reactor Trip Due to High Steam Generator Level

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Regarding Manual Reactor Trip Due to High Steam Generator Level
ML16187A380
Person / Time
Site: Farley Southern Nuclear icon.png
Issue date: 07/05/2016
From: Pierce C
Southern Co, Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-16-0857 LER 16-001-00
Download: ML16187A380 (6)
v • d • e


LER-2016-001, Regarding Manual Reactor Trip Due to High Steam Generator Level
Event date:
Report date:
Reporting criterion: 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)
10 CFR 50.73(a)(2)(x)
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)(i)
3642016001R00 - NRC Website
v • d • e

text

Charles R. Pierce Regulatory Affairs Director July 5, 2016 Docket Nos.: 50-364 Southern Nuclear Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham. AL 35242 Tel 205.992.7872 Fax 205.992.7601 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 SOUTHERN<<\\

NUCLEAR A SOUTHERN COMPANY NL-16-0857 Joseph M. Farley Nuclear Plant-Unit 2 Licensee Event Report 2016-001-00 Manual Reactor Trip due to High Steam Generator Level Ladies and Gentlemen:

This Licensee Event Report is being submitted pursuant to the requirements of the Code of Federal Regulations, 10 CFR 50. 73(a)(2)(iv)(A) for a manual actuation of the Reactor Protection System and an automatic start of the Auxiliary Feedwater system.

This letter contains no NRC commitments. If you have any questions regarding the submittal, please contact Mr. John Mclean at (334) 814-3342.

RZec~lly f);;d*

C. R. Pierce Regulatory Affairs Director CRP/JWM/cg Enclosure: Unit 2 Licensee Event Report 2016-001-00

U. S. Nuclear Regulatory Commission NL-16-0857 Page2 cc:

Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. M. D. Meier, Vice President-Regulatory Affairs Mr. D. R. Madison, Vice President - Fleet Operations Mr. B. J. Adams, Vice President - Engineering Ms. C. A. Gayheart, Vice President-Farley Ms. B. L. Taylor, Regulatory Affairs Manager-Farley Mr. J. E. Purcell, Site Operating Experience Coordinator RTYPE: CFA04.054 U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. S. A. Williams, NRR Project Manager-Farley Mr. P. K. Niebaum, Senior Resident Inspector-Farley

Enclosure Joseph M. Farley Nuclear Plant-Unit 2 Unit 2 Licensee Event Report 2016-001-00 Manual Reactor Trip due to High Steam Generator Level

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION ~PPROVED BY OMB: NO. 3160-0104 EXPIRES: 10/31/2018 (11-2015)

!, the NRC may not conduct or sponsor, and a person is no

~uired to respond to, the Information collection.

3. PAGE Joseph M. Farley Nuclear Plant, Unit 2 05000-I 364 1 OF 2
4. TITLE Manual Reactor Trip due to High Steam Generator Level
5. EVENT DATE
6. LER NUMBER
7. REPORT DATE
8. OTHER FACILITIES INVOLVED I

SEQUENTIAL I REV NO ACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER MONTH DAY YEAR 05000-07 OS 2016 ACIUTYNAME DOCKET NUMBER 05 11 2016 2016 001 -

00 05000-

9. OPERATING MODE
11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 0 20.2201(b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A) 1 0 20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) 0 20.2203(a)(1) 0 20.2203(a)(4) 0 50. 73(a)(2)Qii) 0 50.73(a)(2)Qx)(A) 0 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A) 18J 50. 73(a)(2)(iv)(A) 0 50.73(a)(2)(x)
10. POWER LEVEL 0 20.2203(a)(2)(ii) 0 50.36(c)(1)(ii)(A) 0 50. 73(a)(2)(v)(A) 0 73.71(a)(4) 0 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50. 73(a)(2)(v)(B) 0 73.71 (a)(5) 29%

0 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii) 0 50. 73(a)(2)(v)(C) 0 73.77(a)(1) 0 20.2203(a)(2)(v) 0 50. 73(a)(2)(i)(A) 0 50.73(a)(2)(v)(D) 0 73.77(a)(2)(i) 0 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 50. 73(a)(2)(vii) 0 73.77(a)(2)(ii) 0 50.73(a)(2)(i)(C) 0 OTHER Specify In Abstract below or in =

A. PLANT AND SYSTEM IDENTIFICATION

Westinghouse - Pressurized Water Reactor Energy Industry Identification Codes are identified in the text as [XX].

8. DESCRIPTION OF EVENT

YEAR

3. LER NUMBER SEQUENTIAL NUMBER 2016 001 On 5/11/2016 at 0653 COT, with Unit 2 (U2) at 29 percent power, the Hi-Hi Steam Generator (SG) setpoint was reached. This caused the main feedwater valves to isolate, the running main feedwater pumps to trip, automatic start of the Motor Driven Auxiliary Feed Pumps, and the main turbine to trip automatically. The reactor was manually tripped per procedure.

C. UNIT STATUS AT TIME OF EVENT Unit 2, Mode 1, 29 percent power

D. CAUSE OF EVENT

The cause of the manual reactor trip was determined to be inadequate control of the feedwater system, leading to an overfilling of the Steam Generators.

E.

REPORT ABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable as required by 10 CFR 50. 73(a)(2)(iv)(A) due to a manual actuation of the reactor protection system and automatic actuation of the auxiliary feedwater system. The reactor was shut down at 0653 and mode 3 was entered to complete the necessary procedural actions. The Motor Driven Auxiliary Feed Pumps also started automatically which is reportable by 10 CFR 50. 73(a)(2)(iv)(A). There was no loss of safety function and no radioactive release associated with this event. All required safety systems were available and the plant responded as expected. There were no actual consequences detrimental to the health and safety of the public and is considered to be of very low safety significance.

F.

CORRECTIVE ACTION

Corrective actions included additional training provided to the startup control room team on manipulations that affect the feedwater system. Also, more specific guidance on feedwater system operation and control during Startup from Hot Standby to Power Operations will also be incorporated into operating procedures.

G. ADDITIONAL INFORMATION

Other system affected:

No systems other than those mentioned in this report were affected by this event.

REV NO.

00

Commitment Information:

This report does not create any licensing commitments

Previous Similar Events

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

2. DOCKET NUMBER
3. LER NUMBER YEAR SEQUENTIAL REV NUMBER NO.

05000-364 2016 001 00 LER 2010-002-00: Reactor Trip due to Failed Feedwater Regulating Valve Controller

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