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


==Dear Mr. Browning:==
==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).
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 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available 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,
Sincerely,
/RA/
/RA/
Gregory E. Werner, Chief Project Branch E Division of Reactor Projects Dockets No.:
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
50-313; 50-368 Licenses No.:
 
DRP-51; NPF-6 Enclosure: Inspection Report 05000313/20130 12; 05000 368/2013012 Pages A3-8 and A3-9  
ML14101A219 ML14101A214
 
the failure of once-through cooling. The evaluation of consequential loss of offsite power provided a dominant accident sequence involving a transient with consequential loss of offsite power, the loss of all feedwater to the steam generators and failure of once-through cooling.
 
Table 2 Core Damage Sequences Sequence Description  Point % of Cut Set Estimate Total 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-  2.74E-7 1.0 527 DGR08H-EFW MAN-SGDEPLT 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 main feedwater, loss of auxiliary feedwater, loss of emergency feedwater, and the failure of once-through cooling. The top ten sequence cutsets are provided in Table 2 of the detailed risk evaluation.
 
A3-8
 
The results are dominated by one core damage sequence. The largest contributor is Sequence 14 from the loss of main feedwater tree. The sequence comprises a failure of all feedwater to the steam generators, including main feedwater, auxiliary feedwater, and emergency feedwater, with a loss of once-through cooling. The remainder of the sequences are dominated by failure of the emergency diesel generators without recovery of ac power.
 
(6) Sensitivity Analysis The SRA performed a variety of uncertainty and sensitivity analyses on the internal events model as shown below. The results confirm the recommended Yellow finding.


Electronic Distribution for Arkansas Nuclear One
Sensitivity Analysis 1 - Transient without Loss of Main Feedwater.


A3-8 t he fa ilure of once-through cooling. The evaluation of consequential loss of o ff site pow er provided a domina nt accide nt sequence involvi ng a transien t with consequential loss of o ffsite power, t he loss of all feedwater to t he stea m gener ators and fa ilure o f once-through cooling.
The SRA ran the model using a transient as the initiator. The change in core damage frequency was 1.10 x 10-5 (Yellow).


T able 2 Core D a m a g e S e q u e n c es S e q u e n c e De scr i pt i on P o i nt E s t i mate % of T ot a l Cut S et Cou n t M F W-14 I E M F W-F W-O T C 2.6 9 E-5 95.6 6, 0 3 6 LO O P-19 I E LO O P-E F W-O T C 3.7 9 E-7 1.3 1, 7 3 3 LO O P-20-09-10 I E LO O P-SBO(EP S)-R S U B-O PR08H- D G R08H-E F W M A N-S G D E P LT 2.7 4 E-7 1.0 527 M F W-15-10 I E M F W-R PS-F W A T W S 1.2 5 E-7 0.4 157 M F W-13 I E M F W-F W-SS RC-H P R 8.9 8 E-8 0.3 1, 6 7 9 LO O P-20-30 I E LO O P-SB O-E F W-O P R08H-D GR08H 8.0 0 E-8 0.3 959 M F W-02-09-04 I E M F W-LO S C-RCP T-H P I 6.1 4 E-8 0.2 814 M F W-15-11 I E M F W-R PS-RCS P R E SS U RE 3.9 9 E-8 0.1 18 M F W-15-09 I E M F W-R PS-B O R A T I O N 3.7 9 E-8 0.1 16 M F W-12 I E M F W-F W-SS CR-C S R 2.6 3 E-8 0.1 560 O th e r s A l l A d d i t i o n a l S e q u e n c es C o m b i n e d 1.3 3 E-7 0.5 3, 8 8 6 T ot a l CC D P A l l S e q u e n c es 2.8 1 e-5 1 0 0.0 1 6 , 3 85 A b b r e v i ati o ns B O RA TI O N Fail ure o f E m e r g e n cy Bora t i on CBO C o n tr o lled Bl e e d o f f I so lated CSR C o n t a inment S pray R ec i r cu l ati o n D G R 0 8 H N o n r ec o v ery of Di es e l G e n era t o r i n 8 H o u r s E F W E m e r g e n cy F e e d w ater E F W M A N M a n u a l C o n trol o f E m e r g e ncy F e e d w ater EPS E m e r g e n cy P o w er S y s t em FW F e e d w ater S y s t em (M F W , E F W , a n d a u x ili a r y f e e d w ate r) F W A T W S F e e d w ater S y s t em u n d e r AT W S C o n d i t i o n s HPI Hi gh Pre ssure I n j e c t i on HPR Hi gh Pre ssure R ec i r cu l a t i on I E L O O P I n i t i ati n g E v e n t: L o s s o f O ff s i t e P o w er I E M F W I n i t i ati n g E v e n t: L o s s o f M a i n Fee d w ater LOSC L o ss o f RC P S e a l C ooli ng O PR 0 8 H N o n r ec o v ery of O ff s i t e P o w er i n 8 H o u r s O T C O nce-T h r o u g h C o o li n g RCPT R e a c t or C o ol a n t P u m ps T r i p p ed RCSPRESS RC S P r ess u re L i m i t ed RSUB R e a c t or C o ol a n t S u b co o l i ng M a i nta i n e d RPS R e a c t or P r ot e c t ion S y s t e m SBO S t ati on B l a c k o u t S G DEP LT L a t e D e pr e ssuri z ati o n o f S t e a m G e n e r at o r s SSCR S ec o n dary C ooli ng R ec o v ered T he domina nt accident sequence cutsets involved a loss of mai n feedwater, loss of aux iliary feedwater , loss of emergency feedwater , a n d t he fa ilure of once-through cooling. T he t op t en sequence cutsets are provided in Ta ble 2 of the deta il ed risk evaluation.
Sensitivity Analysis 2 - No consequential loss of offsite power.


A3-9 T he r esults are dominated by one core damag e seque nce. T he largest con tributor is Sequence 14 fr om t he loss o f main feedwater tree. The sequence comprises a fa ilure of all feedwater to the stea m gen erat or s , i nclu di ng main feedwater, aux iliary feedwater , and em ergency feedwater , with a loss of once-through cooling. T he remaind er o f the sequences ar e dominated by fa ilure o f the em ergency di esel gener ators without r ecover y of ac power. (6) Sensitivity An alysis T he SRA performed a variety of uncertainty an d sensitivity analyses on t he i nternal events model as shown below. The r es ults confir m the r ecommended Ye llow fin ding. Sensitivity An alysis 1 Transien t without Lo ss o f Main Feedw ater. T he SRA r an t he mod el us i ng a transient as t he initiator. T he change i n core damage frequency was 1.10 x 10-5 (Ye llow). 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 risk from a postulated consequential loss of offsite power.


T he SRA r an t he mod el witho ut i nclu ding t he 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).


T he change in core damage frequency was 2.
Sensitivity Analysis 3 - Potential Recovery of Bus 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 methodology, represented the probability that operators would fail to recover the bus prior to core damage, given the adverse and unknown conditions of site electrical supply. The change in core damage frequency was 1.97 x 10-5 (Yellow).


74 x 10-5 (Ye llow). Sensitivity An alysis 3 Potential R ecovery of B us 2A2 T he SRA r an t he mod el wi t h t he fa ilure of Bus 2A2 probability set to 6.79 x 10-1. This value, calculated using SPAR-H method ology , repr esent ed the probability that op erators wo uld fai l t o recov er t he bus prio r to core damage , giv en t he adverse and unknown conditions of s it e electrical supply. T he change in core damage frequency was 1.
(7) Contributions from External Events (Fire, Flooding, and Seismic)
Manual Chapter 0609, Appendix A, Section 6.0 requires, when the internal events detailed risk evaluation results are greater than or equal to 1.0E-7, the finding should be evaluated for external event risk contribution. The analyst noted that this detailed risk assessment evaluates an actual event in which no external events occurred.


97 x 10-5 (Ye llow). (7) Con tributions fr o m Externa l Events (Fire, Flooding, and Seismic) Manua l Chapter 0609, Appendix A, Sectio n 6.0 requires, whe n t he internal events deta iled risk evaluati on results are greate r tha n or equal to 1.0E-7, t he fin ding should be evaluat ed f or ext ernal event risk con tribution. T he analyst not ed tha t this deta il ed risk assessment evaluates an actual event i n which no external events occurred.
Additionally, the period of time that the events impacted plant equipment was small enough that the probability of an external initiator occurring during this time would be negligible. Therefore, the analyst assumed that the risk from external events, given the subject performance deficiency was essentially zero.


Ad ditionally, the period o f time that the events impacted plan t equipmen t was small enoug h tha t the probability of an ext ernal initiator occurrin g during this time wo uld be neg ligibl e. Theref or e , the ana lyst assum ed tha t t he risk from external events, giv en the subject perfo rmance deficiency was essenti ally zero.
A3-9
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Latest revision as of 05:50, 4 November 2019

Errata for Arkansas Nuclear One - NRC Augmented Inspection Team Follow-Up Report 05000313/2013012 and 05000368/2013012
ML14101A219
Person / Time
Site: Arkansas Nuclear  Entergy icon.png
Issue date: 04/10/2014
From: Greg Werner
NRC/RGN-IV/DRP/RPB-E
To: Jeremy G. Browning
Entergy Operations
Werner G
References
IR-13-012
Download: ML14101A219 (4)
v • d • e


Text

ril 10, 2014

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 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available 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

ML14101A219 ML14101A214

the failure of once-through cooling. The evaluation of consequential loss of offsite power provided a dominant accident sequence involving a transient with consequential loss of offsite power, the loss of all feedwater to the steam generators and failure of once-through cooling.

Table 2 Core Damage Sequences Sequence Description Point % of Cut Set Estimate Total 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- 2.74E-7 1.0 527 DGR08H-EFW MAN-SGDEPLT 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 main feedwater, loss of auxiliary feedwater, loss of emergency feedwater, and the failure of once-through cooling. The top ten sequence cutsets are provided in Table 2 of the detailed risk evaluation.

A3-8

The results are dominated by one core damage sequence. The largest contributor is Sequence 14 from the loss of main feedwater tree. The sequence comprises a failure of all feedwater to the steam generators, including main feedwater, auxiliary feedwater, and emergency feedwater, with a loss of once-through cooling. The remainder of the sequences are dominated by failure of the emergency diesel generators without recovery of ac power.

(6) Sensitivity Analysis The SRA performed a variety of uncertainty and sensitivity analyses on the internal events model as shown below. The results confirm the recommended Yellow finding.

Sensitivity Analysis 1 - Transient without Loss of Main Feedwater.

The SRA ran the model using a transient as the initiator. The change in core damage frequency was 1.10 x 10-5 (Yellow).

Sensitivity Analysis 2 - No consequential loss of offsite power.

The SRA ran the model without including the additional runs to calculate the change in risk from a postulated consequential loss of offsite power.

The change in core damage frequency was 2.74 x 10-5 (Yellow).

Sensitivity Analysis 3 - Potential Recovery of Bus 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 methodology, represented the probability that operators would fail to recover the bus prior to core damage, given the adverse and unknown conditions of site 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, Section 6.0 requires, when the internal events detailed risk evaluation results are greater than or equal to 1.0E-7, the finding should be evaluated for external event risk contribution. The analyst noted that this detailed risk assessment evaluates an actual event in which no external events occurred.

Additionally, the period of time that the events impacted plant equipment was small enough that the probability of an external initiator occurring during this time would be negligible. Therefore, the analyst assumed that the risk from external events, given the subject performance deficiency was essentially zero.

A3-9

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