05000348/LER-2014-003, Regarding Scaling Errors Result in Inoperable Steam Flow Channels for Durations Longer than Allowed by Technical Specifications
| ML14153A684 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/02/2014 |
| From: | Gayheart C Southern Co, Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-14-0838 LER 14-003-00 | |
| Download: ML14153A684 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3482014003R00 - NRC Website | |
text
Cheryl A.Gayheart SouthernNuclear Vice Presicteni - Farley Operating Company, Inc.
Farley Nuclcar Plant Post Office Drawer 470 Ashford.Alabama 36312 Tel 334.814.4511 Fax 334.814.4575 June 2,2014 COMPANY Docket No.;
50-348 NL-14-0838 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. FarleyNuclearPlant-Units 1 and 2 Licensee Event Report 2014-003-00 Scaling Errors Result in Inoperable SteamFlow Channels for Durations Longer Than Allowed bv Technical Specifications Ladies and Gentlemen:
This Licensee Event Report, "Scaling Errors Result inInoperable SteamFlow Channels for Durations Longer ThanAllowed byTechnical Specifications," is being submitted pursuant tothe requirements of theCode of Federal Regulations, 10 CFR 50.73(a)(2)(i)(B) as an operation orcondition prohibited by Technical Specifications.
This letter contains no NRC commitments. If you have any questions regarding the submittal, please contact Mr. Bill Arens at (334) 814-4765.
Sincerely,
>heart Vice Pre^deVtt - Farley CAG/WNA Enclosure: Units 1 and 2 Licensee Event Report 2014-003-00
U. S. Nuclear Regulatory Commission NL-14-0838 Page 2 cc:
Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President &CEO IVIr. D. G. Bost, Executive Vice President &Chief Nuclear Officer Mr. B. L Ivey, Vice President - Regulatory Affairs Mr. C. R. Pierce, Regulatory Affairs Director Mr. D. R. Madison, Vice President - Fleet Operations Mr. R. R. Martin, Regulatory Affairs Manager - Farley Mr. J. E. Purcell, Nuclear Technical Specialist-Farley Ms. K. A.Walker, Senior Engineer - Farley RTYPE: CFA04.054 U. S. Nuclear Reoulatorv Commission Mr. V. M. McCree, Regional Administrator Mr. S. A. Williams, NRR Project Manager - Farley Mr. P. K. Niebaum, SeniorResidentInspector - Farley Mr. J. R. Sowa, Resident Inspector - Farley Mr. R. E. Martin, Senior Project Manager-Farley
- Insanh M. Farlev Nuclear Plant - Units 1 and 2 NL-14-0838 Scaling Errors Result in Inoperable Steam Flow Channels for Durations Longer Than Allowed by Technical Specifications Enclosure Units 1 and 2 Licensee Event Report 2014-003-00
NRC FORM 366 (02-20t4)
U.S. NUCLEAR REGULATORYCOMMISSION APPROVED BY OMB: NO. 31504104 EXPIRES: 01/31/2017 LICENSEE EVENT REPORT (LER)
(See Page 2 forrequired numberof digits/characters foreach block)
Eslimaled Ifte NRC may not conduct Ofspotw*. andaperson IS not LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET Joseph M. Farley Nuclear Plant, Unit 1 05000 348
NARRATIVE
YEAR 2014
- 6. LER NUMBER SEQUENTIAL t^MBER 003 REV NO.
00
- 3. PAGE 2 of 4 Westinghouse - Pressurized WaterReactor Energy Industry Identification Codes are identified in the text as[XXJ.
Requirement for Report This report is required per 10 CFR 50.73(a)(2)(i)(B) as acondition prohibited by Technical Specifications due to the Unit 1Steam Flow Channel Q1C22FT0484 [FT] and the Unit 2Steam Flow Chanel Q2C22FT0494 [FT] being inoperable without the applicable Required Actions of Techniral Specification 3.3.2 (Engineered Safety Feature Actuation System (ESFAS) Instrumentation) being perfomied within the requiredCompletion Time.
Unit Status at Time of Event From the time of the first discovery on April 2.2014 of an incorrectly scaled steam flow channel until both channels were properly scaled, both units operated in Mode 1at 100 percent thermal power.
Description of Event
At 1819 on April 2,2014, during an extent of condition review of an unrelated scaling issue, Engineering personnel identified that Unit 1procedures for calibration of steam flow channels contained scaling data due to the transmitter span information not being updated from the P^fvious ^erating^cj^^^^
Further review identified that as aresult of the incorrect scaling data, steam flow chann^Q1C22^0484, the IB Steam Generator Channel 111 Steam Flow instrument, was outside of its a"owed Technical Specification tolerance. The steam flow channel was declared inoperable and the applicable Technical Specification Required Action was entered at 2104 on April 2,2014. The steam flow subsequently re-calibrated using acorrected steam flow calibration procedure and
- 0 oP status at 0821 on April 5,2014. Since this steam flow instrument had been incorrectly calibrated during the most recent refueling outage in the Fall of 2013, the 72-hour Required Action o'Jfhnical Specification 3.3.2 (ESFAS Instrumentation) to place the channel in the tripped condition that becarne applicable on October 25,2013 was not met due to station personnel being unaware of The remainder of the Unit 1steam flow channels, although impacted by the scaling error, remained within the allowableTechnical Specificationtolerance.
On April 8,2014, during continued extent of condition reviews. Engineering personnel Identified that Unit 2procedures for calibration of steam flow channels also contained transmitter span data that was ^
updated from the previous cycle. Further review of Unit 2steam generator scaling data idei^ifted that steam flow channel Q2C22FT0494, the 2C Steam Generator Channel III Steam Flow "Jf
^
outside of its allowed Technical Specification tolerance. The steam flow channel was and the applicable Technical Specification Required Action was entered at 1145 on April 8,2014 Th^
steam flow instrument was subsequently re-calibrated using acorrected steam flow P"^^""^
and restored to operable status at 0040 on April 9,2014. Since this steam flow instrument had l^en incorrectly calibrated during the most recent refueling outage in the Spring of 2013, Action of Technical Specification 3.3.2 (ESFAS Instrumentation) to place the channel '"^he tripped condHion that becameapplicable on May8,2013 was not tnetdue to S! haJST the scaling error. The remainder of the Unit 2steam flow channels, although impacted by the scaling error, remained within the allowable Technical Specification tolerance.
NRC FORM 3<<6A (Ol-MM)
lNRCFORM3e6A l(02-:014)
U.S.NUCLEAR REGULATORY COMMISSION LICENEE EVENT REPORT (LER) 1 1.FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE 1
1 Joseph M. Farley Nuclear Plant, Unit 1 05000 348 YEAR SEQUENTIAL NUMBER REV NO.
3of4 1
2014 003 00
- narrative
Cause of Event
The direct cause of this event is due to steam flow instrument transmitter span data tables not being updated with current cycle information prior to instrument calibration during the most recent refueling outage for each unit. These instrument tables had not been updated due to acombination of an oniission in procedure FNP-O-ETP-4509 (Scaling Engineer Guidelines) and knowledge weaknesses among the Engineering personnel performing the scaling activities.
Safety Assessment
Steam flow channels perfonn asafety function by providing a high steam flow input to main steam line isolation logic circuitry. Each of the three steam generators is equipped with two redundant steam flow transmitters. Ahigh steam flow signal from one of the two steam flow transmitters on two of the three steam generators coincident with a low-low reactor coolant system (RCS) average temperature signal from two of three RCS temperature channels generates a main steam line isolation signal that causes closure of all main steam line isolation valves.
During the time periods of the Unit 1FT-484 and Unit 2FT-494 out-of-tolerance condition, the redundant steam flow transmitters remained capable of performing their safety function. Therefore, sufficient inputs to the main steam line isolation circuitry were available to actuate a mam steam line isolation at the proper setpoint. At no time was there a loss of safety function. Additionally, Unit 1FT-484 and Unit 2 FT-494 remained capable of actuating during a high steam flow condition, although slightly above the allowed setpoint.
Adiverse means of providing a main steam line isolation in the event of asteam line break is the low-steam-pressure main steam line isolation signal. This function remained fully capable of performing the main steam line isolation function during the periods that these two steam flow channels were known to be out of tolerance.
Adiverse means of providing a main steam line isolation in the event of asteam line break inside Containment is the high-high Containment pressure main steam line isolation signal. This 'unction remained fully capable of performing the main steam line isolation function during the periods that these twosteam flow channels were known to be out oftolerance.
Steam flow channels also provide acontrol function associated with the steam generator water level control system. The steam flow channel scaling errors caused minimal impact to the steam generator water level control systems with no observable effects on the function of these systems.
Based on the above considerations this condition is considered to have low safety significance.
Corrective Action
Unit 1steam flow transmitter FT-484 was rescaled per work order SNC564824 and returned to operable status at 0821 on April 5,2014. Unit 2steam flow transmitter FT-494 was rescaled per work order SNC566091 and was returned to operable status at0040 on April 9,2014. The remainder of the instruments impacted by this scaling error issue have been rescaled per separate work orders.
Calibration procedures for all protection instrument channels on both uriits controlled by the pr(wedure FNP-O-ETP-4509 (Scaling Engineer Guidelines) have been reviewed with no other instances of scaling errors being identified that impacted channel operability.
NRC FORM 3<<6A (tC-3)l4>
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- 3. PAGE 4 Of 4 The error in procedure FNP-O-ETP-4509 (Scaling Engineer Guidelines) hasbeen corrected. Additionally, animproved and less cumbersome process for incorporating scaling data into calibration procedures is being pursued.
Training of Engineering personnel regarding this event hasbeen conducted. An engineer training plan for scaling and normalization processes will be developed.
Additional Information
Joseph M. Farley Nuclear Plant Unit 2 LER 2013-001-00 wassubmitted onJuly 26,2013to report two instances of a condition prohibited by Technical Specifications duetotheuntimely renormalization of Steam FlowTransmitter FT-494 at the beginning of two operating cycles.
Joseph M. Farley Nuclear Plant Unit 1 LER 2013-003-00 was submitted on January 3,2014toreport a condition prohibited by Technical Specifications duetothe untimely renormalization ofSteam Flow TransmitterFT-495 at the beginning of an operatingcycle.