IR 05000313/2013012

Errata for Arkansas Nuclear One - NRC Augmented Inspection Team Follow-Up Report 05000313/2013012 and 05000368/2013012

ML14101A219
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/10/2014
From:	Werner G E NRC/RGN-IV/DRP/RPB-E
To:	Jeremy G. Browning Entergy Operations
Werner G E
References
IR-13-012
Download: ML14101A219 (4)

Text

April 10, 2014

Jeremy Browning, Site Vice President Entergy Operations, Inc. Arkansas Nuclear One 1448 SR 333 Russellville, AR 72802-0967

SUBJECT: ERRATA FOR ARKANSAS NUCLEAR ONE NRC AUGMENTED INSPECTION TEAM FOLLOW-UP REPORT 05000313/2013012 AND 05000368/2013012

Dear Mr. Browning:

Please remove pages A3-8 and A3-9 from the NRC Inspection Report 05000313/2013012 and 05000368/2013012 and replace them with the pages enclosed with this letter. The purpose of this change is to correct an administrative error in the detailed risk evaluation associated with Unit 2. In accordance with Title 10 of the Code of Federal Regulations ns, s Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the lable Records (PARS) component of the NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/

Gregory E. Werner, Chief Project Branch E Division of Reactor Projects Dockets No.:

50-313; 50-368 Licenses No.:

DRP-51; NPF-6

Enclosure:

Inspection Report 05000313/2013012; 05000368/2013012 Pages A3-8 and A3-9

Electronic Distribution for Arkansas Nuclear One

A3-8 the failure of once-through cooling. The evaluation of consequential loss of offsite power provided a domina nt accident sequence involvi ng a transient with consequential loss of o ffsite power, t he loss of all feedwater to the steam generators and fa ilure of once-through cooling.

Table 2 Core Damage Sequences Sequence Description Point Estimate % of Total Cut Set Count MFW-14 IEMFW-FW-OTC 2.69E-5 95.6 6,036 LOOP-19 IELOOP-EFW-OTC 3.79E-7 1.3 1,733 LOOP-20-09-10 IELOOP-SBO(EPS)-RSUB-OPR08H- DGR08H-EFWMAN-SGDEPLT 2.74E-7 1.0 527 MFW-15-10 IEMFW-RPS-FWATWS 1.25E-7 0.4 157 MFW-13 IEMFW-FW-SSRC-HPR 8.98E-8 0.3 1,679 LOOP-20-30 IELOOP-SBO-EFW-OPR08H-DGR08H 8.00E-8 0.3 959 MFW-02-09-04 IEMFW-LOSC-RCPT-HPI 6.14E-8 0.2 814 MFW-15-11 IEMFW-RPS-RCSPRESSURE 3.99E-8 0.1 18 MFW-15-09 IEMFW-RPS-BORATION 3.79E-8 0.1 16 MFW-12 IEMFW-FW-SSCR-CSR 2.63E-8 0.1 560 Others All Additional Sequences Combined 1.33E-7 0.5 3,886 Total CCDP All Sequences 2.81e-5 100.0 16,385 Abbreviations BORATION Failure of Emergency Boration CBO Controlled Bleedoff Isolated CSR Containment Spray Recirculation DGR08H Nonrecovery of Diesel Generator in 8 Hours EFW Emergency Feedwater EFWMAN Manual Control of Emergency Feedwater EPS Emergency Power System FW Feedwater System (MFW, EFW, and auxiliary feedwater) FWATWS Feedwater System under ATWS Conditions HPI High Pressure Injection HPR High Pressure Recirculation IELOOP Initiating Event: Loss of Offsite Power IEMFW Initiating Event: Loss of Main Feedwater LOSC Loss of RCP Seal Cooling OPR08H Nonrecovery of Offsite Power in 8 Hours OTC Once-Through Cooling RCPT Reactor Coolant Pumps Tripped RCSPRESS RCS Pressure Limited RSUB Reactor Coolant Subcooling Maintained RPS Reactor Protection System SBO Station Blackout SGDEPLT Late Depressurization of Steam Generators SSCR Secondary Cooling Recovered The dominant accident sequence cutsets involved a loss of mai n feedwater, loss of aux iliary feedwater

, loss of emergency feedwater

, and the failure of once-through cooling. T he top ten sequence cutsets are provided in Ta ble 2 of the deta iled risk evaluation.

A3-9 The results are dominated by one core damag e sequence. The largest contributor is Sequence 14 from the loss of main feedwater tree. The sequence comprises a fa ilure of all feedwater to the stea m generators, including main feedwater, aux iliary feedwater

, and emergency feedwater

, with a loss of once-through cooling. T he remainder of the sequences ar e dominated by fa ilure of the emergency diesel generators without r ecovery of ac power. (6) Sensitivity An alysis The SRA performed a variety of uncertainty an d sensitivity analyses on the internal events model as shown below

. The results confir m the recommended Ye llow finding. Sensitivity An alysis 1 Transien t without Lo ss of Main Feedw ater. The SRA ran the model using a transient as t he initiator. T he change i n core damage frequency was 1.10 x 10-5 (Yellow). Sensitivity An alysis 2 No consequentia l loss of offsite power.

The SRA ran the model without including the additional runs to calculate the change in ri sk from a postulated consequentia l loss of o ffsite power.

The change in core damage frequency was 2.

74 x 10-5 (Yellow). Sensitivity An alysis 3 Potential R ecovery of B us 2A2 The SRA ran the model with the failure of Bus 2A2 probability set to 6.79 x 10-1. This value, calculated using SPAR-H method ology, represented the probability that op erators would fail to recover the bus prior to core damage

, given the adverse and unknown conditions of s ite electrical supply

. The change in core damage frequency was 1.

97 x 10-5 (Yellow). (7) Contributions from External Events (Fire, Flooding, and Seismic) Manual Chapter 0609, Appendix A, Sectio n 6.0 requires, when the internal events deta iled risk evaluati on results are greate r than or equal to 1.0E-7, the finding should be evaluat ed for external event risk contribution. The analyst not ed that this deta iled risk assessment evaluates an actual event i n which no external events occurred.

Additionally, the period o f time that the events impacted plan t equipment was small enoug h that the probability of an external initiator occurrin g during this time wo uld be negligible. Therefore, the analyst assumed that the risk from external events, giv en the subject perfo rmance deficiency was essentially zero.