NL-13-1514, LER 13-001-00 for Joseph M. Farley, Unit 2, Regarding 2C Steam Generator Flow Transmitter Inoperable Longer than Allowed by Technical Specifications: Difference between revisions

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{{Adams
#REDIRECT [[05000364/LER-2013-001]]
| number = ML13207A374
| issue date = 07/26/2013
| title = LER 13-001-00 for Joseph M. Farley, Unit 2, Regarding 2C Steam Generator Flow Transmitter Inoperable Longer than Allowed by Technical Specifications
| author name = Lynch T A
| author affiliation = Southern Co, Southern Nuclear Operating Co, Inc
| addressee name =
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000364
| license number =
| contact person =
| case reference number = NL-13-1514
| document report number = LER 13-001-00
| document type = Letter, Licensee Event Report (LER)
| page count = 6
}}
 
=Text=
{{#Wiki_filter:Joseph M. Farley Nuclear Plant - Unit 2 NL-13-1514 2C Steam Generator Flow Transmitter Inoperable Longer Than Allowed By Technical Specifications Unit 2 Licensee Event Report 2013-001-00
 
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION(10-2010)
LICENSEE EVENT REPORT (LER)
APPROVED BY OMB:  NO. 3150-0104 EXPIRES:  10/31/2013 Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the FOIA/Privacy Section (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to infocollects.resource@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 maynot conduct or sponsor, and a person is not required to respond to, the information collection.1. FACILITY NAME Joseph M. Farley Nuclear Plant, Unit 2
: 2. DOCKET NUMBER 05000 364 3. PAGE 1  OF  4 4. TITLE  2C Steam Generator Flow Transmitter Inoperable Longer Than Allowed by Technical Specifications
: 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 29 2013 2013 - 001 - 00 07 26 2013FACILITY NAME      DOCKET NUMBER
: 9. OPERATING MODE 1 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§:  (Check all that apply) 20.2201(b)  20.2203(a)(3)(i)  50.73(a)(2)(i)(C)  50.73(a)(2)(vii)  20.2201(d)  20.2203(a)(3)(ii)  50.73(a)(2)(ii)(A)  50.73(a)(2)(viii)(A)  20.2203(a)(1)  20.2203(a)(4)  50.73(a)(2)(ii)(B)  50.73(a)(2)(viii)(B)  20.2203(a)(2)(i)  50.36(c)(1)(i)(A)  50.73(a)(2)(iii)  50.73(a)(2)(ix)(A)  20.2203(a)(2)(ii)  50.36(c)(1)(ii)(A)  50.73(a)(2)(iv)(A)  50.73(a)(2)(x) 20.2203(a)(2)(iii)  50.36(c)(2)  50.73(a)(2)(v)(A)  73.71(a)(4)  20.2203(a)(2)(iv)  50.46(a)(3)(ii)  50.73(a)(2)(v)(B)  73.71(a)(5)  20.2203(a)(2)(v)  50.73(a)(2)(i)(A)  50.73(a)(2)(v)(C)  OTHER  20.2203(a)(2)(vi)  50.73(a)(2)(i)(B)  50.73(a)(2)(v)(D)        Specify in Abstract below          or in NRC Form 366A 10. POWER LEVEL 100 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME        J.M. Farley Nuclear Plant, William N. Arens - Licensing Supervisor TELEPHONE NUMBER (Include Area Code)
  (334) 814-4765 13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX CAUSE SYSTEMCOMPONENT MANU-FACTURER REPORTABLE TO EPIX X JE FT B080 Y      14. SUPPLEMENTAL REPORT EXPECTED  YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
NO 15. EXPECTED SUBMISSION DATE MONTH DAY YEAR 09 13 2013 ABSTRACT (Limit to 1400 spaces, i.e., approximat ely 15 single-spaced typewritten lines)
On May 29, 2013, with Unit 2 operating in Mode 1 at 100% power, Engineering personnel performing a review of Unit 2 beginning-of-cycle power ascension data identified that 2C Steam Generator Steam Flow Transmitter FT-494 did not meet Technical Specification calibration accuracy requirements. Based on this information the steam flow instrument was declared inoperable and the required actions of the appropriate Technical Specification were performed. However, since the data utilized in the engineering review was obtained on May 14, 2013, it is known that FT-494 has been inoperable since May 14, 2013. Consequently, the time limits of the applicable Technical Specification required action were not met. This represents a condition prohibited by Technical Specifications and is reportable under 10CFR50.73(a)(2)(i)(B). Steam flow transmitter FT-494 was re-calibrated and returned to service on June 1, 2013. Investigation into the cause of the FT-494 inaccuracy has not been completed. This out-of-tolerance condition of FT-494 also occurred at the beginning of the previous fuel cycle.
NRC FORM 366A (10-2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION1. FACILITY NAME 2. DOCKET  6. LER NUMBER 3. PAGE Joseph M. Farley Nuclear Plant, Unit 2 05000 364 YEAR SEQUENTIAL NUMBER REVISION NUMBER 2 of 4 2013 -001 - 00 NARRATIVE    NRC FORM 366A (10-2010)
Westinghouse - Pressurized Water Reactor Energy Industry Identification Codes are identified in the text as [XX].
 
Description of Event
 
Farley Nuclear Plant steam flow transmitters are calibrated each refueling outage utilizing
 
instrument data from the previous operating cycle. Following the initial post-refueling ascension in power the steam flow transmitters are normalized utilizing beginning-of-cycle data per procedure FNP-2-ETP-4462 (Power Ascension Following a Unit Refueling Outage).
On May 29, 2013 at 1339, with Unit 2 operating in Mode 1 at 100% power, Engineering personnel performing beginning-of-cycle normalization calculations determined that 2C Steam Generator Steam Flow Transmitter FT-494 [FT] was outside of Technical Specification accuracy requirements. Technical Specifications require that the high steam flow bistable actuates at less than 110.3% of full steam flow. However, the power ascension data indicated that the actual high steam flow setpoint for FT-494 was 111.38% of full steam flow.
 
As a result of the determination by Engineering, FT-494 was declared inoperable and the Technical Specification 3.3.2 (Engineered Safety Feature Actuation System Instrumentation) required action for the inoperable channel was entered. The effected bistable was placed in the tripped condition within the required 72 hours from the declaration of inoperability. FT-494 was subsequently re-calibrated and the high steam flow bistable was verified to trip at a setpoint less than the Technical Specification limit of 110.3% of full steam flow. FT-494 was returned to operable status at 1642 on June 1, 2013. FT-494 is a model number 764 flow transmitter
 
manufactured by ITT Barton.
The data utilized in the Engineering determination was obtained on May 14, 2013. For the purpose of determining reportability, this date is considered the point of discovery.
Consequently, the completion times associated with the applicable Technical Specification required actions were not met. This represents an operation or condition prohibited by Technical Specifications and is reportable under 10CFR50.73(a)(2)(i)(B).
 
The Engineering review also identified that a si milar condition existed at the beginning of the previous fuel cycle. At 1311 on January 4, 2012, with Unit 2 operating in Mode 1 at 100%
power, Engineering review of beginning-of-cycle instrument data identified that the FT-494 high steam flow bistable setpoint was 111.36% of full steam flow which is above the maximum Technical Specification allowable setpoint of 110.3% of full steam flow. This resulted in a declaration of inoperability for FT-494. FT-494 was recalibrated and returned to service at 0935 on January 7, 2012 which is within the time requirement of the applicable Technical Specification required action based on the time of declaration of inoperability. The data utilized for this review was obtained on November 29, 2011, which is the discovery time for the purpose of determining reportability. Based on this discovery time, the completion time associated with the applicable Technical Specification required action was not met. This also represents an Operation or Condition Prohibited by Technical Specifications and is reportable under NRC FORM 366A (10-2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION1. FACILITY NAME 2. DOCKET  6. LER NUMBER 3. PAGE Joseph M. Farley Nuclear Plant, Unit 2 05000 364 YEAR SEQUENTIAL NUMBER REVISION NUMBER 3 of 4 2013 -001 - 00 NARRATIVE    NRC FORM 366A (10-2010) 10CFR50.73(a)(2)(i)(B). Due to an oversight by the plant staff this reportability requirement was not recognized and the required Licensee Event Report was not submitted.
 
Cause of Event
 
Causal analysis of the out-of-tolerance condition identified on January 4, 2012 was performed utilizing the Apparent Cause Determination process. This causal analysis determined that the out of tolerance condition could be due to transmitter issues such as the equalizing valve leak-by or partial clogging of the flow transmitter sensing lines. This causal analysis included a review of associated procedures and work order packages for calculation or implementation errors with no errors being identified. As a result of this analysis, actions were put in place to monitor the performance of FT-494 over a six-month period, collect and review data to resolve possible sensing line blockage, and to investigate the condition of the transmitter equalizing valve during the Spring 2013 Unit 2 refueling outage. The causal analysis also specified performance of a re-evaluation of the event causes following the completion of the above mentioned actions. This re-analysis is being completed with the Enhanced Apparent Cause
 
Determination for the May 29, 2013 event.
 
The out-of-tolerance condition identified on May 29, 2013, is being evaluated using the Enhanced Apparent Cause Determination process. This cause determination is being conducted in conjunction with the re-analysis of the previous causal determination and is scheduled for completion on July 26, 2013. This Licensee Event Report will be supplemented upon the completion of these causal analyses.
 
Safety Assessment Steam Flow Transmitter FT-494 performs a safety 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. A high 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 FT-494 out-of-tolerance condition, the redundant steam flow transmitter remained capable of performing its safety function. Therefore, sufficient inputs to the main steam line isolation circuitry were available to actuate a main steam line isolation at the proper setpoint. At no time was there a loss of safety function.
 
A diverse means of providing a main steam line isolation in the event of a steam 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 FT-494 was known to be out of tolerance.
 
NRC FORM 366A (10-2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION1. FACILITY NAME 2. DOCKET  6. LER NUMBER 3. PAGE Joseph M. Farley Nuclear Plant, Unit 2 05000 364 YEAR SEQUENTIAL NUMBER REVISION NUMBER 4 of 4 2013 -001 - 00 NARRATIVE    NRC FORM 366A (10-2010)
Based on the above considerations and on the FT-494 setpoint being slightly above allowable tolerance but otherwise operable this condition is considered to have low safety significance.
 
Corrective Action
 
FT-494 was rescaled per work order SNC 459224 and returned to service at 1642 on June 1, 2013. Further corrective actions for the FT-494 high steam flow setpoint being above the allowable value have yet to be determined. Upon completion of the causal analysis, this
 
Licensee Event Report will be supplemented by September 13, 2013.
The failure of the plant staff to identify the event that occurred on January 4, 2012 as requiring a Licensee Event Report has been entered in the plant corrective action program as condition
 
report 675409.
 
Additional Information
 
A review of previously submitted Licensee Event Reports identified no similar issues being reported within the previous three years. There have been three other instances in the previous three years of steam flow transmitters requiring re-normalization following the evaluation of beginning-of-cycle data. In all of these cases, the high steam flow setpoints remained within the accuracy requirements of Technical Specifications and operability of the steam flow channel was not affected. These occurrences are documented in condition reports 55575 (for Unit 1 FT-
 
485), 485511 (for Unit 1 FT-484), and 646824 (for Unit 2 FT-485).}}

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