IR 05000327/2014004: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 October 30, 2014  
[[Issue date::October 30, 2014]]


Mr. Joseph Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801
Mr. Joseph Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801
Line 32: Line 31:
If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Sequoyah Nuclear Plant  
If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Sequoyah Nuclear Plant  


If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the Sequoyah Nuclear Plant. In accordance with Title 10 of the Code of Federal Regulations 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC'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 NRC's Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the Sequoyah Nuclear Plant. In accordance with Title 10 of the Code of Federal Regulations 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC'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 NRC's Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Managem ent System (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).


Sincerely,/RA/  
Sincerely,/RA/  


Jonathan H. Bartley, Chief Reactor Projects Branch 6 Division of Reactor Projects Docket Nos.: 50-327, 50-328 License Nos.: DPR-77, DPR-79  
Jonathan H. Bartley, Chief Reactor Projects Branch 6  
 
Division of Reactor Projects Docket Nos.: 50-327, 50-328 License Nos.: DPR-77, DPR-79  


===Enclosure:===
===Enclosure:===
Line 46: Line 47:
_________________________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRP RII:DRP RII:DRP RII:DRS RII:DRP SIGNATURE Via email Via email JHB /RA for/ Via email JHB /RA/ NAME GSmith WDeschaine CKontz PBraaten JBartley DATE 10/29/2014 10/29/2014 10/29/2014 10/28/2014 10/30/2014 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO Letter to Joseph from Jonathan H. Bartley dated October 30, 2014
_________________________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRP RII:DRP RII:DRP RII:DRS RII:DRP SIGNATURE Via email Via email JHB /RA for/ Via email JHB /RA/ NAME GSmith WDeschaine CKontz PBraaten JBartley DATE 10/29/2014 10/29/2014 10/29/2014 10/28/2014 10/30/2014 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO Letter to Joseph from Jonathan H. Bartley dated October 30, 2014


SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000327/2014004 AND 05000328/2014004 DISTRIBUTION: D. Gamberoni, RII L. Gibson, RII OE Mail RIDSNRRDIRS PUBLIC R. Pascarelli, NRR RidsNrrPMSequoyah Resource Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket Nos.: 50-327, 50-328 License Nos.: DPR-77, DPR-79 Report Nos.: 05000327/2014004, 05000328/2014004 Licensee: Tennessee Valley Authority (TVA)
SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000327/2014004 AND 05000328/2014004  
 
DISTRIBUTION
: D. Gamberoni, RII L. Gibson, RII OE Mail RIDSNRRDIRS  
 
PUBLIC R. Pascarelli, NRR RidsNrrPMSequoyah Resource  
 
Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket Nos.: 50-327, 50-328  
 
License Nos.: DPR-77, DPR-79  
 
Report Nos.: 05000327/2014004, 05000328/2014004  
 
Licensee: Tennessee Valley Authority (TVA)  
 
Facility: Sequoyah Nuclear Plant, Units 1 and 2 Location: Sequoyah Access Road Soddy-Daisy, TN 37379  
Facility: Sequoyah Nuclear Plant, Units 1 and 2 Location: Sequoyah Access Road Soddy-Daisy, TN 37379  


Dates: July 1 - September 30, 2014 Inspectors: G .Smith, Senior Resident Inspector W. Deschaine, Resident Inspector C. Kontz, Senior Project Inspector P. Braaten, Project Inspector Approved by: Jonathan H. Bartley, Chief Reactor Projects Branch 6 Division of Reactor Projects Enclosure  
Dates: July 1 - September 30, 2014  
 
Inspectors: G .Smith, Senior Resident Inspector W. Deschaine, Resident Inspector C. Kontz, Senior Project Inspector P. Braaten, Project Inspector Approved by: Jonathan H. Bartley, Chief Reactor Projects Branch 6 Division of Reactor Projects  
 
Enclosure  


=SUMMARY=
=SUMMARY=
Line 63: Line 82:
Specifically, during a refueling outage on May 24, 2014, the licensee failed to properly install a locking wire associated with the spring tension bolts on the Unit 2 containment vacuum relief valve. This error ultimately led to a failure of the valve on June 24 at 1600 and entry into TS 3.6.3, "Containment Isolation Valves."  The valve was ultimately repaired and the valve was declared operable on June 26 at 0026. The inspectors determined that the licensee's failure to adequately develop and implement a procedure governing the maintenance of a containment isolation valve was a performance deficiency.
Specifically, during a refueling outage on May 24, 2014, the licensee failed to properly install a locking wire associated with the spring tension bolts on the Unit 2 containment vacuum relief valve. This error ultimately led to a failure of the valve on June 24 at 1600 and entry into TS 3.6.3, "Containment Isolation Valves."  The valve was ultimately repaired and the valve was declared operable on June 26 at 0026. The inspectors determined that the licensee's failure to adequately develop and implement a procedure governing the maintenance of a containment isolation valve was a performance deficiency.


This finding was determined to be greater than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstone's objective to ensure the structural integrity of the containment boundary. Specifically, the finding challenged containment integrity. A screening analysis was conducted using the assumption that all core damage sequences would lead to a Large Early Release. This was an overestimation of risk, since actions to mitigate a release were possible. The short exposure time multiplied by the Core Damage Frequency for the plant resulted in less than a 1E-7 increase in Large Early Release Probability, and the finding is
This finding was determined to be greater than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstone's objective to ensure the structural integrity of the containment boundary. Specifically, the fi nding challenged containment integrity. A screening analysis was conducted using the assumption that all core damage sequences would lead to a Large Early Release. This was an overestimation of risk, since actions to mitigate a release were possible. The short exposure time multiplied by the Core Damage Frequency for the plant resulted in less than a 1E-7 increase in Large Early Release Probability, and the finding is
: '''Green.'''
: '''Green.'''
The cause of this finding was determined to have a cross-cutting aspect in the Human Performance component, relating to the assurance by supervision that procedures are adequate to ensure nuclear safety. [H.1].  (Section 1R12)
The cause of this finding was determined to have a cross-cutting aspect in the Human Performance component, relating to the assurance by supervision that procedures are adequate to ensure nuclear safety. [H.1].  (Section 1R12)  


===B. Licensee-Identified Violations===
===B. Licensee-Identified Violations===
Violations of very low safety significance, which were identified by the licensee were reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program (CAP). These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.
Violations of very low safety significance, which were identified by the licensee were reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program (CAP). These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.


=REPORT DETAILS=
=REPORT DETAILS=


Summary of Plant Status:
===Summary of Plant Status===
Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection period. Unit 2 operated at or near 100 RTP for the entire inspection period.
:
Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection  
 
period. Unit 2 operated at or near 100 RTP for the entire inspection period.


==REACTOR SAFETY==
==REACTOR SAFETY==
Line 89: Line 112:
No findings were identified.
No findings were identified.


===2. Summer Readiness of Offsite and Alternate AC Power Systems:===
===2. Summer Readiness of Offsite and Alternate AC Power Systems===
:


====a. Inspection Scope====
====a. Inspection Scope====
Line 170: Line 194:
* documentation of activities, including initials and sign-offs in procedures
* documentation of activities, including initials and sign-offs in procedures
* supervision of activities, including risk and reactivity management
* supervision of activities, including risk and reactivity management
* pre-job briefs Specifically, the inspectors observed licensed operator performance during the following activities:
* pre-job briefs Specifically, the inspectors observed licensed operator performance during the following activities:
* reactivity manipulation involving the dilution of boric acid in the reactor coolant system (RCS)
* reactivity manipulation involving the dilution of boric acid in the reactor coolant system (RCS)
* reactivity manipulation involving the addition of boric acid in the RCS Documents reviewed are listed in the Attachment. This activity constituted one inspection sample.
* reactivity manipulation involving the addition of boric acid in the RCS Documents reviewed are listed in the Attachment. This activity constituted one inspection sample.


====b. Findings====
====b. Findings====
Line 193: Line 217:
On June 24 at 1600, the main control began a routine purge of the containment atmosphere. The purge was secured at 1647. At 2125, Operations personnel noted the vacuum relief valve, 2-VLV-30-573 was not in its normally closed position. This was evidenced by two of the three limit switches indicated the valve was partially open. Following a review of the issue, operations declared the containment isolation function of the vacuum breaker to be inoperable at 2315 and entered TS LCO 3.6.3 action 'a' which required the penetration to be isolated in 4 hours. Maintenance personnel were dispatched to the valve and noted that indeed the valve was cracked open. Maintenance personnel adjusted the valve and established adequate seating and action 'a' was exited at 0311 on June 25. However, the vacuum breaker was deemed inoperable and thus LCO 3.6.6 for the vacuum breaker was entered. This LCO had an action time of 72 hours. VLV-30-573 was subsequently repaired, retested and declared back in service at 0026 on June 26.
On June 24 at 1600, the main control began a routine purge of the containment atmosphere. The purge was secured at 1647. At 2125, Operations personnel noted the vacuum relief valve, 2-VLV-30-573 was not in its normally closed position. This was evidenced by two of the three limit switches indicated the valve was partially open. Following a review of the issue, operations declared the containment isolation function of the vacuum breaker to be inoperable at 2315 and entered TS LCO 3.6.3 action 'a' which required the penetration to be isolated in 4 hours. Maintenance personnel were dispatched to the valve and noted that indeed the valve was cracked open. Maintenance personnel adjusted the valve and established adequate seating and action 'a' was exited at 0311 on June 25. However, the vacuum breaker was deemed inoperable and thus LCO 3.6.6 for the vacuum breaker was entered. This LCO had an action time of 72 hours. VLV-30-573 was subsequently repaired, retested and declared back in service at 0026 on June 26.


An apparent cause evaluation revealed that a locking device (a lock wire) was not appropriately installed on the spring tension bolts inside the valve on May 24 during the last Unit 2 refueling outage. Instead, the lock wire was misapplied to the valve body. This ultimately led to a slow degradation of the valve as the spring nut was allowed to gradually back out and thus reduce the closing spring pressure on the valve. During repair of the valve, it was noted that the as-found spring pressure on the valve was approximately 17 psi while the as-left was nominally 30 psi. The perturbation of the containment purge activity lowered containment pressure enough to momentarily actuate the valve open and the reduced closing spring force allowed the valve to become misaligned as it reclosed. At this point, the valve was partially open and considered failed as it could no longer perform its containment isolation function of closing.
An apparent cause evaluation revealed that a locking device (a lock wire) was not appropriately installed on the spring tension bolts inside the valve on May 24 during the last Unit 2 refueling outage. Instead, the lock wire was misapplied to the valve body. This ultimately led to a slow degradation of the valve as the spring nut was allowed to gradually back out and thus reduce the closing spring pressure on the valve. During repair of the valve, it was noted that the as-found spring pressure on the valve was approximately 17 psi while the as-left was nominally 30 psi. The perturbation of the  
 
containment purge activity lowered containment pressure enough to momentarily actuate the valve open and the reduced closing spring force allowed the valve to become misaligned as it reclosed. At this point, the valve was partially open and considered failed as it could no longer perform its containment isolation function of closing.


=====Analysis.=====
=====Analysis.=====
The licensee's failure to adequately develop and implement a procedure governing the maintenance of a containment isolation valve was a performance deficiency. Specifically, failure to follow a work order package precluded the proper installation of a lock wire inside a containment isolation valve. This then led to the gradual loosening of the spring nut which led to a reduction of the closing spring force on the valve. The containment purge evolution led to the ultimate failure of the valve during the momentary opening and closing of the valve. The inspectors evaluated this issue in accordance with the NRC's significance determination process (SDP). This finding was determined to be greater than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstone's objective to ensure the structural integrity of the containment boundary. Specifically, the finding challenged containment integrity. Using IMC 0609.04, Initial Characterization of Findings and IMC 0609 Appendix A, Exhibit 3 - Barrier Screening Questions, the finding required analysis using MC 0609, Appendix 'H', "Containment Integrity Significance Determination Process," as the finding represented an open pathway in the physical integrity of containment via a valve. According to Appendix H, the finding represented a Type B finding as it did not directly affect the core damage frequency. The finding was then processed under Section 6 using a phase 2 analysis. The inspectors performed the phase 2 analysis using Sec 6.2 of Appendix H. The inspectors determined, based on the actual circumstances being evaluated, that a more detailed assessment needed to be performed in a SDP Phase 3 evaluation. The regional Senior Reactor Analyst performed a Phase 3 SDP for the finding. A screening analysis was conducted using the assumption that all core damage sequences would lead to a Large Early Release. This was an overestimation of risk, since actions to mitigate a release were possible.
The licensee's failure to adequately develop and implement a procedure governing the maintenance of a containment isolation valve was a performance deficiency. Specifically, failure to follow a work order package precluded the proper installation of a lock wire inside a containment isolation valve. This then led to the gradual loosening of the spring nut which led to a reduction of the closing spring force on the valve. The containment purge evolution led to the ultimate failure of the valve during the momentary opening and closing of the valve. The inspectors evaluated this issue in accordance with the NRC's significance determination process (SDP). This finding was determined to be greater than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstone's objective to ensure the structural integrity of the containment boundary. Specifically, the finding challenged containment integrity. Using IMC 0609.04, Initial Characterization of Findings and IMC 0609 Appendix A, Exhibit 3 - Barrier Screening Questions, the finding required analysis using MC 0609, Appendix 'H', "Containment Integrity Significance Determination Process," as the finding represented an open pathway in the physical integrity of containment via a valve. According to Appendix H, the finding represented a Type B finding as it did not directly affect the core damage frequency. The finding was then processed under Section 6 using a phase 2 analysis. The inspectors performed the phase 2 analysis using Sec 6.2 of Appendix H. The inspectors determined, based on the actual circumstances being evaluated, that a more detailed assessment needed to be performed in a SDP Phase 3 evaluation. The regional Senior Reactor Analyst performed a Phase 3 SDP for the finding. A screening analysis was conducted using the assumption that all core damage sequences would lead to a Large Early Release. This was an overestimation of risk, since actions to mitigate a release were possible.


The short exposure time multiplied by the Core Damage Frequency for the plant resulted in less than a 1E-7 increase in Large Early Release Probability, and the finding is GREEN. The cause of this finding was determined to have a cross-cutting aspect in the Human Performance component, relating to the assurance by supervision that procedures are adequate to ensure nuclear safety. [H.1].
The short exposure time multiplied by the Core Damage Frequency for the plant resulted in less than a 1E-7 increase in Large Early Release Probability, and the finding is GREEN. The cause of this finding was determined to have a cross-cutting aspect in the Human Performance component, relating to the assurance by supervision that procedures are adequate to ensure nuclear safety. [H.1].  


=====Enforcement:=====
=====Enforcement:=====
Unit 2 TS 6.8.1.a. required, in part, that written procedures be established, implemented, and maintained covering the activities specified in Appendix A, "Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors," of Regulatory Guide (RG) 1.33, "Quality Assurance Program Requirements (Operations)," Revision 2, dated February 1978. RG 1.33 Appendix A Section 9.a, "Procedures for Performing Maintenance," required, in part, that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. The work instructions from WO 07-779292-000 (step 5.7.21) required installation of the safety wire (lock wire) but did not provide any detail for how this was to be accomplished. Contrary to the above, on May 24, 2014, the licensee failed to properly install the safety wire such that the spring nut was precluded from loosening. Specifically, the maintenance procedure failed to provide adequate detail and guidance to ensure the safety wire was properly installed to ensure the spring nut was adequately locked. Corrective actions included:  1) repair of the vacuum relief valve;    2) the development of a maintenance case study for inclusion into maintenance training; and 3) Revision of the operating instruction to provide validation of containment vacuum relief valve position following containment purges. Because the finding was of very low safety significance and has been entered into the licensee's CAP as PER 902721, this violation is being treated as an NCV, consistent with the NRC Enforcement Policy:  NCV 05000328/2014004-01, Failure of Containment Vacuum Relief.
Unit 2 TS 6.8.1.a. required, in part, that written procedures be established, implemented, and maintained covering the activities specified in Appendix A, "Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors," of Regulatory Guide (RG) 1.33, "Quality Assurance Program Requirements (Operations)," Revision 2, dated February 1978. RG 1.33 Appendix A Section 9.a, "Procedures for Performing Maintenance," required, in part, that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. The work instructions from WO 07-779292-000 (step 5.7.21) required installation of the safety wire (lock wire) but did not provide any detail for how this was to be accomplished. Contrary to the above, on May 24, 2014, the licensee failed to properly install the safety wire such that the spring nut was precluded from loosening. Specifically, the maintenance procedure failed to provide adequate detail and guidance to ensure the safety wire was properly installed to ensure the spring nut was adequately locked. Corrective actions included:  1) repair of the vacuum relief valve;    2) the development of a maintenance case study for inclusion into maintenance training; and 3) Revision of the operating instruction to provide validation of containment vacuum relief valve position following containment purges. Because the finding was of very low safety significance and has been entered into the licensee's CAP as PER 902721, this violation is being treated as an NCV, consiste nt with the NRC Enforcement Policy:  NCV 05000328/2014004-01, Failure of Containment Vacuum Relief.
{{a|1R13}}
{{a|1R13}}
==1R13 Maintenance Risk Assessments and Emergent Work Control==
==1R13 Maintenance Risk Assessments and Emergent Work Control==


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the following activities to determine whether appropriate risk assessments were performed prior to removing equipment from service for maintenance. The inspectors evaluated whether risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent work was performed, the inspectors reviewed whether plant risk was promptly reassessed and managed. The inspectors also assessed whether the licensee's risk assessment tool use and risk categories were in accordance with Standard Programs and Processes Procedure NPG-SPP-07.1, "On-Line Work Management," Revision 3, and Instruction 0-TI-DSM-000-007.1, "Risk Assessment Guidelines," Revision 9. The inspectors completed five samples.
The inspectors reviewed the following activities to determine whether appropriate risk assessments were performed prior to removing equipment from service for maintenance. The inspectors evaluated w hether risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent work was performed, the inspectors reviewed whether plant risk was promptly reassessed and managed. The inspectors also assessed whether the licensee's risk assessment tool use and risk categories were in accordance with Standard Programs and Processes Procedure NPG-SPP-07.1, "On-Line Work Management," Revision 3, and Instruction 0-TI-DSM-000-007.1, "Risk Assessment Guidelines," Revision 9. The inspectors completed five samples.
* emergent work 2A EDG Lube oil recirculation pump replacement
* emergent work 2A EDG Lube oil recirculation pump replacement
* emergent failure of overcurrent 51 relay on 1B CHS pump breaker
* emergent failure of overcurrent 51 relay on 1B CHS pump breaker
Line 219: Line 245:


====a. Inspection Scope====
====a. Inspection Scope====
For the operability evaluations described in the problem event reports (PERs) listed below, the inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available, such that no unrecognized increase in risk occurred. The inspectors compared the operability evaluations to the Updated Final Safety Analysis Report (UFSAR) descriptions to determine if the system or component's intended function(s) were adversely impacted. In addition, the inspectors reviewed compensatory measures implemented to determine whether the compensatory measures worked as stated and the measures were adequately controlled. The inspectors also reviewed a sampling of PERs to assess whether the licensee was identifying and correcting any deficiencies associated with operability evaluations. The inspectors completed four samples.
For the operability evaluations described in the problem event reports (PERs) listed below, the inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and t he subject component or system remained available, such that no unrecognized increase in risk occurred. The inspectors compared the operability evaluations to the Updated Final Safety Analysis Report (UFSAR) descriptions to determine if the system or component's intended function(s) were adversely impacted. In addition, the inspectors reviewed compensatory measures implemented to determine whether the compensatory measures worked as stated and the measures were adequately controlled. The inspectors also reviewed a sampling of PERs to assess whether the licensee was identifying and correcting any deficiencies associated with operability evaluations. The inspectors completed four samples.
* PER 907134, prompt determination of operability (PDO) on  2A Pipe Chase Cooler leak
* PER 907134, prompt determination of operability (PDO) on  2A Pipe Chase Cooler leak
* PER 922622, PDO on Unit 1 Ice Condenser Icing
* PER 922622, PDO on Unit 1 Ice Condenser Icing
Line 234: Line 260:
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the temporary modification listed below and the associated 10 CFR 50.59 screening, and compared it against the UFSAR and TS to verify whether the modification affected operability or availability of the affected system.
The inspectors reviewed the temporary modification listed below and the associated 10 CFR 50.59 screening, and compared it against the UFSAR and TS to verify whether the modification affected operability or availability of the affected system.
* TACF - SQN-1-2014-085-001 - DC Wetting Current Source Installation for Rod Position Indicator (RPI) E-5 Following installation and testing, the inspectors observed indications affected by the modification, discussed them with operators, and verified that the modification was installed properly and its operation did not adversely affect safety system functions.
* TACF - SQN-1-2014-085-001 - DC Wetting Current Source Installation for Rod Position Indicator (RPI) E-5 Following installation and testing, the inspectors observed indications affected by the modification, discussed them with operators, and verified that the modification was installed properly and its operation did not adversely affect safety system functions.


The inspectors completed one sample.
The inspectors completed one sample.
Line 257: Line 283:


====a. Inspection Scope====
====a. Inspection Scope====
Unit 2 Refueling Outage Heat-up and Start-up Activities To verify reactor coolant safety (RCS) integrity and containment integrity, the inspectors further reviewed the licensee's RCS leakage calculations and containment isolation valve lineups. In order to verify that core operating limit parameters were consistent with core design, the inspectors also observed portions of the low power physics testing, including reactor criticality. The inspectors also reviewed the physics testing results. The inspectors completed one sample.
Unit 2 Refueling Outage Heat-up and Start-up Activities
 
To verify reactor coolant safety (RCS) integrity and containment integrity, the inspectors further reviewed the licensee's RCS leakage calculations and containment isolation valve lineups. In order to verify that core operating limit parameters were consistent with core design, the inspectors also observed portions of the low power physics testing, including reactor criticality. The inspectors also reviewed the physics testing results. The inspectors completed one sample.


====b. Findings====
====b. Findings====
Line 267: Line 295:
For the surveillance tests identified below, the inspectors assessed whether the structures, systems, and components (SSCs) involved in these tests satisfied the requirements described in the TS surveillance requirements, the UFSAR, applicable licensee procedures, and whether the tests demonstrated that the SSCs were capable of performing their intended safety functions. This was accomplished by witnessing testing and/or reviewing the test data. The inspectors completed three samples.
For the surveillance tests identified below, the inspectors assessed whether the structures, systems, and components (SSCs) involved in these tests satisfied the requirements described in the TS surveillance requirements, the UFSAR, applicable licensee procedures, and whether the tests demonstrated that the SSCs were capable of performing their intended safety functions. This was accomplished by witnessing testing and/or reviewing the test data. The inspectors completed three samples.


Routine Surveillance Tests:
Routine Surveillance Tests
:
* 2-SI-EDC-082-307.A, Undervoltage/Degraded Voltage EDG Start and Load Shedding Time Response Relay Test, Revision 17
* 2-SI-EDC-082-307.A, Undervoltage/Degraded Voltage EDG Start and Load Shedding Time Response Relay Test, Revision 17
* 1-SI-OPS-082-024.A, 1A-A DG 24 hour Run and Load Rejection Test, Rev. 26 RCS leakage test:
* 1-SI-OPS-082-024.A, 1A-A DG 24 hour Run and Load Rejection Test, Rev. 26 RCS leakage test
:
* 0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory, Rev. 34
* 0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory, Rev. 34


Line 276: Line 306:


===Cornerstone:===
===Cornerstone:===
Emergency Preparedness 1EP6 Drill Evaluation
Emergency Preparedness 1EP6 Drill Evaluation


====a. Inspection Scope====
====a. Inspection Scope====
Line 287: Line 317:


==OTHER ACTIVITIES==
==OTHER ACTIVITIES==
Cornerstones:  Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security
Cornerstones:  Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security


==OTHER ACTIVITIES==
==OTHER ACTIVITIES==
Line 294: Line 324:


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors sampled licensee submittals for the Performance Indicators (PI) listed below for the period from July 1, 2013, through June 30, 2014 for both Unit 1 and Unit 2. Definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Indicator Guideline, Revision 6, were used to determine the reporting basis for each data element in order to verify the accuracy of the PI data reported during that period.
The inspectors sampled licensee submittals for the Performance Indicators (PI) listed below for the period from July 1, 2013, through June 30, 2014 for both Unit 1 and Unit 2. Definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Indicator Guideline, Revision 6, were used to determine the reporting basis for each data element in order to verify the accuracy of the PI data reported during that  
 
period.


===Cornerstone:===
===Cornerstone:===
Line 303: Line 335:
* Mitigating Systems Performance Index: Residual Heat Removal System
* Mitigating Systems Performance Index: Residual Heat Removal System
* Mitigating Systems Performance Index: Cooling Water System
* Mitigating Systems Performance Index: Cooling Water System
* Safety System Functional Failures   The inspectors reviewed portions of the operations logs and raw performance index (PI)data developed from monthly operating reports and discussed the methods for compiling and reporting the PIs with engineering personnel. The inspectors also independently calculated selected reported values to verify their accuracy and compared graphical representations from the most recent PI report to the raw data to verify that the data was correctly reflected in the report. Specifically for the Mitigating Systems Performance Index (MSPI), the inspectors reviewed the basis document and derivation reports to verify that the licensee was properly entering the raw data as suggested by NEI 99-02.
* Safety System Functional Failures The inspectors reviewed portions of the operations logs and raw performance index (PI)data developed from monthly operating reports and discussed the methods for compiling and reporting the PIs with engineering personnel. The inspectors also independently calculated selected reported values to verify their accuracy and compared graphical representations from the most recent PI report to the raw data to verify that the data was correctly reflected in the report. Specifically for the Mitigating Systems Performance Index (MSPI), the inspectors reviewed the basis document and derivation reports to verify that the licensee was properly entering the raw data as suggested by NEI 99-02.


For Safety System Functional Failures, the inspectors also reviewed licensee event reports (LERs) issued during the referenced timeframe. The inspectors completed 12 samples (six per unit).
For Safety System Functional Failures, the inspectors also reviewed licensee event reports (LERs) issued during the referenced timeframe. The inspectors completed 12 samples (six per unit).
Line 325: Line 357:


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed an in-depth review of PER 876781. In April 2014, the licensee noted there were four NRC-issued NCVs received within the previous 12 months that exhibited the cross cutting aspect of "evaluation" as described in NRC manual chapter (MC) 0310, "Aspects Within Cross Cutting Areas."  The cross cutting aspects were part of four separate issues (NCVs):  1) Failure to evaluate motor operated valves with regard to lower degraded voltage values, 2) Failure to correct in a timely manner a dual indication on residual heat removal (RHR) valve, 1-FCV-63-72, 3) Failure to install sprinklers in accordance with the fire protection program (FPP) and code requirements, and 4) Failure to perform the required review when adding fire watches to the FPP. MC 0310 describes a cross cutting aspect as a performance characteristic of a finding that is the most significant causal factor of the performance deficiency. Specifically, the four findings above were related to "evaluations" (Section 06.02.P.2 under MC 0310) and noted to contain licensee weaknesses regarding the thorough evaluation of issues to ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance. As a result of this identified weakness, the licensee developed an Apparent Cause Evaluation (ACE). The inspectors reviewed the actions taken to determine if the licensee had adequately addressed the following attributes:
The inspectors performed an in-depth review of PER 876781. In April 2014, the licensee noted there were four NRC-issued NCVs received within the previous 12 months that exhibited the cross cutting aspect of "eval uation" as described in NRC manual chapter (MC) 0310, "Aspects Within Cross Cutting Areas."  The cross cutting aspects were part of four separate issues (NCVs):  1) Failure to evaluate motor operated valves with regard to lower degraded voltage values, 2) Failure to correct in a timely manner a dual indication on residual heat removal (RHR) valve, 1-FCV-63-72, 3) Failure to install sprinklers in accordance with the fire protection program (FPP) and code requirements, and 4) Failure to perform the required review when adding fire watches to the FPP. MC 0310 describes a cross cutting aspect as a performance characteristic of a finding that is the most significant causal factor of the performance deficiency. Specifically, the four findings above were related to "evaluations" (Section 06.02.P.2 under MC 0310) and noted to contain licensee weaknesses regarding the thorough evaluation of issues to ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance.
 
As a result of this identified weaknes s, the licensee developed an Apparent Cause Evaluation (ACE). The inspectors reviewed the actions taken to determine if the licensee had adequately addressed the following attributes:
* complete, accurate and timely identification of the problem
* complete, accurate and timely identification of the problem
* evaluation and disposition of operability and reportability issues
* evaluation and disposition of operability and reportability issues
Line 334: Line 368:


====b. Findings and Observations====
====b. Findings and Observations====
The inspectors noted the ACE to be thorough and probing. The conclusion of the ACE was that the staff had knowledge deficiencies in 1) documenting critical thinking; 2)understanding design and licensing bases; and 3) identifying potential impacts to nuclear safety and compliance when performing required problem evaluations. The corrective actions included training and stand-downs to stress the importance of thoroughly evaluating significant safety and regulatory issues. The inspectors also noted that the licensee performed a detailed extent of condition as well as an extent of cause analysis as part of this ACE.
The inspectors noted the ACE to be thorough and probing. The conclusion of the ACE was that the staff had knowledge deficiencies in 1) documenting critical thinking; 2)understanding design and licensing bases; and 3) identifying potential impacts to nuclear  
 
safety and compliance when performing required problem evaluations. The corrective actions included training and stand-downs to stress the importance of thoroughly evaluating significant safety and regulatory issues. The inspectors also noted that the licensee performed a detailed extent of condition as well as an extent of cause analysis  
 
as part of this ACE.


===.3 Selected Issue Follow-up:===
===.3 Selected Issue Follow-up:===
Line 354: Line 392:
==4OA3 Event Follow-up==
==4OA3 Event Follow-up==


===.1 (Closed) Licensee Event Report (LER) 05000328/2014-002-00, Containment Vacuum Relief Valve Inoperable Resulting in Condition Prohibited by Technical Specifications===
===.1 (Closed) Licensee Event Report (L===
 
ER) 05000328/2014-002-00, Containment Vacuum Relief Valve Inoperable Resulting in Condition Prohibited by Technical Specifications


====a. Inspection Scope====
====a. Inspection Scope====
On June 24, 2014, at 2125 a MCR operator noted that the containment vacuum relief valve 2-VLV-30-573 was not in its normally closed position. The initial characterization of this deficiency was that the position indication was faulty but the valve was still operable. This caused the crew to enter TS LCO 3.6.3 action 'c' (Similarly, LCO 3.6.6). Both LCOs were 72 hours and provided the crew time to deal with the issue. However, based on further review of the issue and telephone conversations with management, the crew concluded that, indeed, the valve was inoperable at 2315. This was based on the fact that a containment purge had been performed earlier in the day and was concluded at 1647. The licensee concluded that based on past experience, a vacuum relief can partially open due to localized low pressure areas that may result when purging the containment. However, it was surmised that in this instance, the vacuum relief failed to properly reseat after momentarily opening as evidenced by one of the three closing limit switched failing to actuate. This defeated the containment isolation function of the penetration itself and thus required entry into LCO 3.6.3 action 'a' which is a four hour LCO to isolate the penetration. The crew ultimately isolated the penetration at 0247 on June 25 by jacking closed and de-powering the vacuum relief isolation valve. The vacuum relief was ultimately repaired and placed back in to service at 0256 on June 26.
On June 24, 2014, at 2125 a MCR operator noted that the containment vacuum relief valve 2-VLV-30-573 was not in its normally closed position. The initial characterization of this deficiency was that the position indication was faulty but the valve was still operable. This caused the crew to enter TS LCO 3.6.3 action 'c' (Similarly, LCO 3.6.6). Both LCOs were 72 hours and provided the crew time to deal with the issue. However, based on further review of the issue and telephone conversations with management, the crew concluded that, indeed, the valve was inoperable at 2315. This was based on the fact that a containment purge had been performed earlier in the day and was concluded at 1647. The licensee concluded that based on past experience, a vacuum relief can partially open due to localized low pressure areas that may result when purging the containment. However, it was surmised that in this instance, the vacuum relief failed to properly reseat after momentarily opening as evidenced by one of the three closing limit switched failing to actuate. This defeated the containment isolation function of the penetration itself and thus required entry into LCO 3.6.3 action 'a' which is a four hour LCO to isolate the penetration. The crew ultimately isolated the penetration at 0247 on June 25 by jacking closed and de-powering the vacuum relief isolation valve. The vacuum relief was ultimately repaired and placed back in to service at 0256 on June 26.


Further review of this event by the licensee, noted that the conditions were readily available at 2125 that the vacuum relief was partially open and the containment penetration was inoperable. Additionally, the licensee concluded that most likely the vacuum relief became inoperable at 1600 when the purge was started. However, the penetration was not isolated until 0311 the following day for a total of slightly more than 11 hours of inoperability. Considering the LCO action time of four hours to isolate the penetration along with the six hours action time, the total time to be in mode 3 for LCO 3.6.3 action 'a' is 10 hours. Thus, in this instance, failing to isolate the penetration in four hours or placing the Unit 2 in mode 3 in the following six hours (10 hours total) was a condition prohibited by TS.
Further review of this event by the licensee, noted that the conditions were readily available at 2125 that the vacuum relie f was partially open and the containment penetration was inoperable. Additionally, the licensee concluded that most likely the vacuum relief became inoperable at 1600 when the purge was started. However, the penetration was not isolated until 0311 the following day for a total of slightly more than 11 hours of inoperability. Considering the LCO action time of four hours to isolate the penetration along with the six hours action time, the total time to be in mode 3 for LCO 3.6.3 action 'a' is 10 hours. Thus, in this instance, failing to isolate the penetration in four hours or placing the Unit 2 in mode 3 in the following six hours (10 hours total) was a condition prohibited by TS.


Thus, this LER was required pursuant to 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by plant technical specifications.
Thus, this LER was required pursuant to 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by plant technical specifications.
Line 369: Line 409:


====a. Inspection Scope====
====a. Inspection Scope====
On December 19, 2013, an operator trainee noted a lack of procedural guidance to perform TS Surveillance Requirement (SR) 4.7.3.b which required the testing of the safety injection system (SIS) start signal to the C-S CCS pump every 18 months. This pump is normally powered from the 480 volt 'A' train safety-related bus on Unit 2 and alternatively can be powered from the 480 volt 'B' train safety-related bus on Unit 1.
On December 19, 2013, an operator trainee noted a lack of procedural guidance to perform TS Surveillance Requirement (SR) 4.7.3.b which required the testing of the  
 
safety injection system (SIS) start signal to the C-S CCS pump every 18 months. This pump is normally powered from the 480 volt 'A' train safety-related bus on Unit 2 and alternatively can be powered from the 480 volt 'B' train safety-related bus on Unit 1.


Based on design documents, the SIS signal from either unit should give a start signal to the C-S CCS pump. Since the pump is normally powered from Unit 2, the SIS start signal from Unit 2 was tested every 18 months per procedure 2-SI-OPS-082-026.B, "Loss of Offsite Power with Safety Injection - D/G 2B-B Test."  This issue was entered into the CAP as PER 826482.
Based on design documents, the SIS signal from either unit should give a start signal to the C-S CCS pump. Since the pump is normally powered from Unit 2, the SIS start signal from Unit 2 was tested every 18 months per procedure 2-SI-OPS-082-026.B, "Loss of Offsite Power with Safety Injection - D/G 2B-B Test."  This issue was entered into the CAP as PER 826482.


An initial investigation into this issue determined the lack of testing to be a missed surveillance. Operations invoked TS SR 4.0.3 as the total functionally required by SR 4.7.3.b was not fully tested. Since the surveillance frequency of SR4.7.3.b was more than 24 hours (18 months) and the expected retest was to be in excess of 24 hours from the time of discovery, the licensee performed a risk evaluation on how to manage the increased risk of an untested component. On January 13, 2014, the performance of SIS start testing from both units to the C-S CCS pump was completed and SR 4.7.3.b was then met. Subsequent investigation completed on January 30, revealed that, in fact, the C-S pump was never tested in the Unit 1 configuration and thus the deficiency was reclassified as a never performed surveillance. Thus, this LER was required pursuant to 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by plant technical specifications as SR 4.7.3.b was deemed never to have been performed. The inspectors discussed the event with operations, maintenance, engineering, and licensee management personnel to gain an understanding of the conditions leading up to the event and assess licensee actions taken following the event. Additionally, the inspectors reviewed the apparent cause evaluation report to assess the detail and thoroughness of the evaluation and the adequacy of the proposed corrective actions.
An initial investigation into this issue determined the lack of testing to be a missed surveillance. Operations invoked TS SR 4.0.3 as the total functionally required by SR 4.7.3.b was not fully tested. Since the surveillance frequency of SR4.7.3.b was more than 24 hours (18 months) and the expected retest was to be in excess of 24 hours from the time of discovery, the licensee performed a risk evaluation on how to manage the increased risk of an untested component. On January 13, 2014, the performance of SIS start testing from both units to the C-S CCS pump was completed and SR 4.7.3.b was  
 
then met.
 
Subsequent investigation completed on January 30, revealed that, in fact, the C-S pump was never tested in the Unit 1 configuration and thus the deficiency was reclassified as a never performed surveillance. Thus, this LER was required pursuant to 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by plant technical specifications as SR 4.7.3.b was deemed never to have been performed.
 
The inspectors discussed the event with operations, maintenance, engineering, and licensee management personnel to gain an understanding of the conditions leading up to the event and assess licensee actions taken following the event. Additionally, the inspectors reviewed the apparent cause evaluation report to assess the detail and thoroughness of the evaluation and the adequacy of the proposed corrective actions.


====b. Findings====
====b. Findings====
Line 381: Line 429:


====a. Inspection Scope====
====a. Inspection Scope====
On June 6, 2014, the licensee completed a review of administrative control of penetrations used for ice condenser maintenance during fuel movement. The licensee concluded that they had been inappropriately crediting engineering evaluations as justification to utilize open penetrations during fuel movement. In 2000, the licensee obtained TS change 99-15 to TS 3.9.4 that permitted some containment penetrations to remain open during fuel movement. Prior to 2000, the licensee used engineering safety evaluation to justify open penetrations during fuel movement. However, in 1999, the NRC deemed the practice of using 10 CFR 50.59 (via engineering evaluation) as inappropriately affected TS 3.9.4 without prior NRC approval and the practice was ceased. Due to a failure to correct a plant procedure 1,2-SI-OPS-088-006.0, "Containment Building Ventilation Isolation," the practice was inadvertently continued from 2000 to 2014 with respect to four containment penetrations: 1) X79A (transfer ice to ice condenser); X79B (transfer ice to ice condenser); X54 (steam generator sludge lancing) and X117 (ice condenser condensate drain). This LER was submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by technical specification. The inspectors discussed the event with operations, maintenance, engineering, and licensee management personnel to gain an understanding of the conditions leading to the event and assess licensee actions taken following the event. Additionally, the inspectors reviewed the apparent cause evaluation report to assess the detail and thoroughness of the evaluation and the adequacy of the proposed corrective actions.
On June 6, 2014, the licensee completed a review of administrative control of penetrations used for ice condenser maintenance during fuel movement. The licensee concluded that they had been inappropriately crediting engineering evaluations as justification to utilize open penetrations during fuel movement. In 2000, the licensee obtained TS change 99-15 to TS 3.9.4 that permitted some containment penetrations to remain open during fuel movement. Prior to 2000, the licensee used engineering safety evaluation to justify open penetrations during fuel movement. However, in 1999, the NRC deemed the practice of using 10 CFR 50.59 (via engineering evaluation) as inappropriately affected TS 3.9.4 without prior NRC approval and the practice was ceased. Due to a failure to correct a plant procedure 1,2-SI-OPS-088-006.0, "Containment Building Ventilation Isolation," the practice was inadvertently continued from 2000 to 2014 with respect to four containment penetrations: 1) X79A (transfer ice to ice condenser); X79B (transfer ice to ice condenser); X54 (steam generator sludge lancing) and X117 (ice condenser condensate drain). This LER was submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by technical specification.


One of the corrective actions was to upgrade plant procedures to ensure strict control of penetration during fuel movement and the elimination of the engineering evaluation avenue to comply with TS 3.9.4. The inspectors reviewed the LER and PER 886970 to verify that the cause of the breeched penetration was identified and whether corrective actions were appropriate.
The inspectors discussed the event with operations, maintenance, engineering, and licensee management personnel to gain an understanding of the conditions leading to the event and assess licensee actions taken following the event. Additionally, the inspectors reviewed the apparent cause evaluation report to assess the detail and thoroughness of the evaluation and the adequacy of the proposed corrective actions.
 
One of the corrective actions was to upgrade plant procedures to ensure strict control of penetration during fuel movement and the elimination of the engineering evaluation avenue to comply with TS 3.9.4.
 
The inspectors reviewed the LER and PER 886970 to verify that the cause of the breeched penetration was identified and whether corrective actions were appropriate.


The licensee's ACE identified that the controlling procedure did not reference that the four affected penetrations were under administrative control during fuel movement. However, there were other procedures in place that in all likelihood would have caused personnel to isolate these penetrations given a fuel handling accident. Notwithstanding the above, strict TS 3.9.4 compliance was not met regarding the above penetrations.
The licensee's ACE identified that the controlling procedure did not reference that the four affected penetrations were under administrative control during fuel movement. However, there were other procedures in place that in all likelihood would have caused personnel to isolate these penetrations given a fuel handling accident. Notwithstanding the above, strict TS 3.9.4 compliance was not met regarding the above penetrations.
Line 396: Line 448:


====a. Inspection Scope====
====a. Inspection Scope====
(Closed) VIO 05000327,328/2012005-05, Failure to Adequately Evaluate and Qualify Molded Case Circuit Breakers   The inspectors reviewed various documents associated with the above violation. This violation dealt with the seismic mounting characteristics of Heinemann circuit breakers which are used in 120volt vital AC applications. The violation discussed several discrepancies between the physical mounting of these circuit breakers in safety related cabinets and the design documents. The violation also questioned the seismic adequacy of the current mounting design. The licensee analyzed the issue in PER 668367 and formalized a consistent mounting design using Micarta board to ensure adequate seating force for the circuit breakers. The licensee tested the new configuration in a laboratory environment to verify the design could withstand an operating basis earthquake as well as a safe shutdown earthquake. The licensee then performed a plant modification, DCN 23189, "Improvements to Design and Documentation for 120 VAC Vital Instrument Power Board Heinemann Breaker Mountings," Rev A. This modification covered 48 circuit breakers on both units and was completed on Unit 1 during the fall 2013 and Unit 2 during the spring 2014. The inspectors reviewed this modification as well as the associated work orders.
(Closed) VIO 05000327,328/2012005-05, Failure to Adequately Evaluate and Qualify Molded Case Circuit Breakers The inspectors reviewed various documents associated with the above violation. This violation dealt with the seismic mounting characteristics of Heinemann circuit breakers which are used in 120volt vital AC applications. The violation discussed several discrepancies between the physical mounting of these circuit breakers in safety related cabinets and the design documents. The violation also questioned the seismic adequacy of the current mounting design. The licensee analyzed the issue in PER 668367 and formalized a consistent mounting design using Micarta board to ensure adequate seating force for the circuit breakers. The licensee tested the new configuration in a laboratory environment to verify the design could withstand an operating basis earthquake as well as a safe shutdown earthquake. The licensee then performed a plant modification, DCN 23189, "Improvements to Design and Documentation for 120 VAC Vital Instrument Power Board Heinemann Breaker Mountings," Rev A. This modification covered 48 circuit breakers on both units and was completed on Unit 1 during the fall 2013 and Unit 2 during the spring 2014. The inspectors reviewed this modification as well as the associated work orders.


====b. Findings====
====b. Findings====
Line 405: Line 457:


===.1 Exit Meeting Summary===
===.1 Exit Meeting Summary===
On October 2, 2014, the resident inspectors presented the inspection results to Mr. Carlin and other members of his staff, who acknowledged the findings. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.  
 
On October 2, 2014, the resident inspectors presented the inspection results to Mr. Carlin and other members of his staff, who acknowledged the findings. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.
 
{{a|4OA7}}
{{a|4OA7}}
==4OA7 Licensee-identified Violations==
==4OA7 Licensee-identified Violations==
Line 442: Line 496:
: [[contact::P. Pratt]], Plant Maintenance Manager  
: [[contact::P. Pratt]], Plant Maintenance Manager  
: [[contact::M. Purcell]], Quality Assurance Manager  
: [[contact::M. Purcell]], Quality Assurance Manager  
: [[contact::P. Simmons]], Plant Manager  
: [[contact::P. Simmons]], Plant Manager  
 
===NRC personnel===
===NRC personnel===
: [[contact::A. Hon]], Project Manager, Office of Nuclear Reactor Regulation  
: [[contact::A. Hon]], Project Manager, Office of Nuclear Reactor Regulation  
Line 449: Line 504:


===Opened and Closed===
===Opened and Closed===
: 05000328/2014004-01 NCV Failure to Perform Adequate Maintenance on Containment Vacuum Relief Valve (Section 1R12)       
: 05000328/2014004-01 NCV Failure to Perform Adequate Maintenance
on Containment Vacuum Relief Valve  
(Section 1R12)       


===Closed===
===Closed===
: 05000327,328/2012005-05 VIO Failure to Adequately Evaluate and Qualify Molded Case Circuit Breakers (Section 4OA5)
: 05000327,328/2012005-05 VIO Failure to Adequately Evaluate and Qualify  
: Molded Case Circuit Breakers (Section  
: 4OA5)
: [[Closes LER::05000327/LER-2014-001]]-00  LER  Never Performed TS Surveillance for C-S   
: [[Closes LER::05000327/LER-2014-001]]-00  LER  Never Performed TS Surveillance for C-S   
: CCS Pump (Section 4OA3)
: CCS Pump (Section 4OA3)
: [[Closes LER::05000328/LER-2014-002]]-00 LER Containment Vacuum Relief Valve Inoperable Resulting in Condition Prohibited
: [[Closes LER::05000328/LER-2014-002]]-00 LER Containment Vacuum Relief Valve Inoperable Resulting in Condition Prohibited  
 
by TS (Section 4OA3)
by TS (Section 4OA3)
: [[Closes LER::05000327/LER-2014-002]]-00  LER  Lack of Administrative Controls for Some        Containment Penetrations During Fuel      Movement Results in Condition Prohibited
: [[Closes LER::05000327/LER-2014-002]]-00  LER  Lack of Administrative Controls for Some        Containment Penetrations During Fuel      Movement Results in Condition Prohibited
Line 467: Line 527:
: 0-PI-OPS-510-001.0, Flood Preparation Equipment Inventory, Revision 8  
: 0-PI-OPS-510-001.0, Flood Preparation Equipment Inventory, Revision 8  
: Section R05:
: Section R05:
: Fire Protection Procedures
: Fire Protection  
===Procedures===
: FPDP-1, Conduct of Fire Protection, Revision 2  
: FPDP-1, Conduct of Fire Protection, Revision 2  
: 0-PI-FPU-317-299.W, Att. 8, Shift Check List, Revision 32
: 0-PI-FPU-317-299.W, Att. 8, Shift Check List, Revision 32
Line 474: Line 535:
: SQN-FPR-Part-II, SQN Fire Protection Report Part II - Fire Protection Plan, Revision 28  
: SQN-FPR-Part-II, SQN Fire Protection Report Part II - Fire Protection Plan, Revision 28  
: Section R06:
: Section R06:
: Flood Protection Measures Work Orders
: Flood Protection Measures  
: WO 11108121224, Check Standing Water Level in Manholes/Handholes Other documents TVA letter to NRC dated May 4, 2007. TVA response to
===Work Orders===
: WO 11108121224, Check Standing Water Level in Manholes/Handholes  
: Other documents
: TVA letter to NRC dated May 4, 2007. TVA response to
: GL 2007-01  
: GL 2007-01  
: Section R11:
: Section R11:
: Licensed Operator Requalification Procedures
: Licensed Operator Requalification  
===Procedures===
: NPG-SPP-17.8.1 rev 10, 11 Licensed Operator Requalification Examination Development and Implementation
: NPG-SPP-17.8.1 rev 10, 11 Licensed Operator Requalification Examination Development and Implementation
: NPG-SPP-22.207 rev 0 Procedure Use and Adherence  
: NPG-SPP-22.207 rev 0 Procedure Use and Adherence  
Line 484: Line 549:
: PER 907424
: PER 907424
: PER 909616  
: PER 909616  
: Other documents 14-3 CPE Crew documentation and remediations OPL273E1403 R1 & 2
: Other documents
: WCAP-17711-NP1 Pressurized Water Reactor Owners Group Westinghouse Emergency Response Guideline Revision 2-Based Critical Tasks Section R12:
: 14-3 CPE Crew documentation and remediations OPL273E1403 R1 & 2
: Maintenance Effectiveness Procedures
: WCAP-17711-NP1 Pressurized Water Reactor Owners Group Westinghouse Emergency Response Guideline Revision 2-Based Critical Tasks  
: Section R12:
: Maintenance Effectiveness  
===Procedures===
: TI-4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -  
: TI-4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -  
: 10CFR50.65, Revision 23  
: 10CFR50.65, Revision 23  
: Other documents CDE #2756 - Failure of Containment Vacuum breaker 2-VLV-30-573   
: Other documents
: CDE #2756 - Failure of Containment Vacuum breaker 2-VLV-30-573   
: Attachment Section R13:
: Attachment Section R13:
: Maintenance Risk Assessments and Emergent Work Evaluation Procedures 0-TI-DSM-000-007.1, Risk Assessment Guidelines, Revision 9
: Maintenance Risk Assessments and Emergent Work Evaluation  
===Procedures===
: 0-TI-DSM-000-007.1, Risk Assessment Guidelines, Revision 9
: NPG-SPP-07.3, Work Activity Risk Management Process, Revision 3
: NPG-SPP-07.3, Work Activity Risk Management Process, Revision 3
: NPG-SPP-07.2.4, Forced Outage or Short Duration Planned Outage Management, Revision 0
: NPG-SPP-07.2.4, Forced Outage or Short Duration Planned Outage Management, Revision 0
: NPG-SPP-07.2, Outage Management, Revision 0
: NPG-SPP-07.2, Outage Management, Revision 0
: GOI-6, Apparatus Operations, Revision 142
: GOI-6, Apparatus Operations, Revision 142  
: Section R15:
: Section R15:
: Operability Evaluations Procedures
: Operability Evaluations  
===Procedures===
: NEDP-22, Functional Evaluations, Rev. 9
: NEDP-22, Functional Evaluations, Rev. 9
: OPDP-8, Limiting Conditions for Operation Tracking, Rev. 5
: OPDP-8, Limiting Conditions for Operation Tracking, Rev. 5
: NPG-SPP-03.5, Regulatory Reporting Requirements, Revision 2
: NPG-SPP-03.5, Regulatory Reporting Requirements, Revision 2  
: PER/SRs
: PER/SRs
: PER 907134, prompt determination of operability (PDO) on 2A Pipe Chase Cooler leak
: PER 907134, prompt determination of operability (PDO) on 2A Pipe Chase Cooler leak
Line 506: Line 578:
: PER 914789, Foreign material (10" rubber material) found in main turbine oil strainer SRs
: PER 914789, Foreign material (10" rubber material) found in main turbine oil strainer SRs
: 932083 &
: 932083 &
: 932087 associated with U2 Eagle 21 TSP failure
: 932087 associated with U2 Eagle 21 TSP failure Section R18:
: Section R18:
: Plant Modifications  
: Plant Modifications Procedures
===Procedures===
: NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Revision 4
: NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Revision 4
: NPG-SPP-09.4, 10
: NPG-SPP-09.4, 10
Line 514: Line 586:
: NPG-SPP-09.5, Temporary Alterations, Revision 0  
: NPG-SPP-09.5, Temporary Alterations, Revision 0  
: Other TACF -
: Other TACF -
: SQN-1-2014-085-001 - DC Wetting Current Source Installation for Rod Position Indicator (RPI) E-5
: SQN-1-2014-085-001 - DC Wetting Current Source Installation for Rod Position Indicator (RPI) E-5  
: Section R19:
: Section R19:
: Post Maintenance Testing Procedures
: Post Maintenance Testing  
===Procedures===
: MMDP-1, Maintenance Management System, Revision 20
: MMDP-1, Maintenance Management System, Revision 20
: MMDP-3, Guidelines for Planning and Execution of Troubleshooting Activities, Revision 6
: MMDP-3, Guidelines for Planning and Execution of Troubleshooting Activities, Revision 6
Line 532: Line 605:
: 116156226, U2 Stroking valves for PMT or release from HO
: 116156226, U2 Stroking valves for PMT or release from HO
: 115725379, RHR HX Outlet valve Attachment Section R20:
: 115725379, RHR HX Outlet valve Attachment Section R20:
: Refueling and Outage Activities Procedures
: Refueling and Outage Activities  
===Procedures===
: FHI-3, Movement of Fuel, Revision 65  
: FHI-3, Movement of Fuel, Revision 65  
: 0-GO-15, Containment Closure Control, Revision 34  
: 0-GO-15, Containment Closure Control, Revision 34  
: 0-GO-13, Reactor Coolant System Drain and Fill Operations, Revision 71
: 0-GO-13, Reactor Coolant System Drain and Fill Operations, Revision 71
: NPG-SPP-08.1, Nuclear Fuel Management, Revision 00  
: NPG-SPP-08.1, Nuclear Fuel Management, Revision 00  
: 0-PI-OPS-000-011.0, "Containment Access Control During Modes 1-4, Revision 1
: 0-PI-OPS-000-011.0, "Containment Access Control During Modes 1-4, Revision 1  
: Section R22:
: Section R22:
: Surveillance Testing Procedures 2-SI-EDC-082-307.A, Undervoltage/Degraded Voltage EDG Start and Load Shedding Time Response Relay Test, Revision 17 1-SI-OPS-082-024.A, 1A-A
: Surveillance Testing  
===Procedures===
: 2-SI-EDC-082-307.A, Undervoltage/Degraded Voltage EDG Start and Load Shedding Time  
: Response Relay Test, Revision 17 1-SI-OPS-082-024.A, 1A-A
: DG 24 hour Run and Load Rejection Test, Revision 26 0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory, Revision 34
: DG 24 hour Run and Load Rejection Test, Revision 26 0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory, Revision 34


==Section 4OA1: Performance Indicator Verification Procedures==
==Section 4OA1: Performance Indicator Verification==
 
===Procedures===
: NPG-SPP-02.2, Performance Indicator Program, Revision 2
: NPG-SPP-02.2, Performance Indicator Program, Revision 2
: NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6
: NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6

Revision as of 08:23, 1 July 2018

IR 05000327/2014004 and 05000328/2014004, July 1, 2014 - September 30, 2014, Sequoyah, Integrated Inspection
ML14303A237
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 10/30/2014
From: Bartley J H
Reactor Projects Region 2 Branch 6
To: Shea J W
Tennessee Valley Authority
References
IR 2014004
Download: ML14303A237 (29)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 October 30, 2014

Mr. Joseph Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801

SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000327/2014004 AND 05000328/2014004

Dear Mr. Shea:

On September 30, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Sequoyah Nuclear Plant, Units 1 and 2. On October 2, the NRC inspectors discussed the results of this inspection with Mr. Carlin and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented one finding of very low safety significance (Green) in this report. This finding was determined to involve a violation of NRC requirements. Additionally, inspectors documented two licensee-identified violations which were determined to be of very low safety significance in this report. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Sequoyah Nuclear Plant

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the Sequoyah Nuclear Plant. In accordance with Title 10 of the Code of Federal Regulations 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC'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 NRC's Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Managem ent System (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/

Jonathan H. Bartley, Chief Reactor Projects Branch 6

Division of Reactor Projects Docket Nos.: 50-327, 50-328 License Nos.: DPR-77, DPR-79

Enclosure:

Inspection Report 05000327/2014004, 05000328/2014004

w/Attachment:

Supplementary Information cc distribution via ListServ

_________________________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRP RII:DRP RII:DRP RII:DRS RII:DRP SIGNATURE Via email Via email JHB /RA for/ Via email JHB /RA/ NAME GSmith WDeschaine CKontz PBraaten JBartley DATE 10/29/2014 10/29/2014 10/29/2014 10/28/2014 10/30/2014 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO Letter to Joseph from Jonathan H. Bartley dated October 30, 2014

SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000327/2014004 AND 05000328/2014004

DISTRIBUTION

D. Gamberoni, RII L. Gibson, RII OE Mail RIDSNRRDIRS

PUBLIC R. Pascarelli, NRR RidsNrrPMSequoyah Resource

Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket Nos.: 50-327, 50-328

License Nos.: DPR-77, DPR-79

Report Nos.: 05000327/2014004, 05000328/2014004

Licensee: Tennessee Valley Authority (TVA)

Facility: Sequoyah Nuclear Plant, Units 1 and 2 Location: Sequoyah Access Road Soddy-Daisy, TN 37379

Dates: July 1 - September 30, 2014

Inspectors: G .Smith, Senior Resident Inspector W. Deschaine, Resident Inspector C. Kontz, Senior Project Inspector P. Braaten, Project Inspector Approved by: Jonathan H. Bartley, Chief Reactor Projects Branch 6 Division of Reactor Projects

Enclosure

SUMMARY

IR 05000327/2014-004, 05000328/2014-004; 7/1-9/30/2014; Sequoyah Nuclear Plant, Units 1 and 2; Event Follow-up

The report covered a three-month period of inspection by resident inspectors and announced inspections by regional inspectors. One violation was identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," (SDP) dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, "Aspects Within Cross Cutting Areas," dated December 19, 2013. All violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy, dated July 9, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process" Revision 5.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Barrier Integrity

  • Green: A Self-revealing Green Non-Cited Violation (NCV) of Technical Specification (TS) 6.8.1.a. was identified for the licensee's failure to adequately implement a maintenance procedure associated with a vacuum relief containment isolation valve.

Specifically, during a refueling outage on May 24, 2014, the licensee failed to properly install a locking wire associated with the spring tension bolts on the Unit 2 containment vacuum relief valve. This error ultimately led to a failure of the valve on June 24 at 1600 and entry into TS 3.6.3, "Containment Isolation Valves." The valve was ultimately repaired and the valve was declared operable on June 26 at 0026. The inspectors determined that the licensee's failure to adequately develop and implement a procedure governing the maintenance of a containment isolation valve was a performance deficiency.

This finding was determined to be greater than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstone's objective to ensure the structural integrity of the containment boundary. Specifically, the fi nding challenged containment integrity. A screening analysis was conducted using the assumption that all core damage sequences would lead to a Large Early Release. This was an overestimation of risk, since actions to mitigate a release were possible. The short exposure time multiplied by the Core Damage Frequency for the plant resulted in less than a 1E-7 increase in Large Early Release Probability, and the finding is

Green.

The cause of this finding was determined to have a cross-cutting aspect in the Human Performance component, relating to the assurance by supervision that procedures are adequate to ensure nuclear safety. [H.1]. (Section 1R12)

B. Licensee-Identified Violations

Violations of very low safety significance, which were identified by the licensee were reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program (CAP). These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection

period. Unit 2 operated at or near 100 RTP for the entire inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

.1 External Flooding Review

a. Inspection Scope

The inspectors reviewed the licensee's readiness prior to the onset of adverse weather that poses a risk of flooding. Specifically, the inspectors reviewed flood design documents and abnormal operating procedure (AOP)-N.03, Flooding. The inspectors walked down flood protection barriers in the auxiliary building and emergency diesel generator (EDG) building and verified required temporary spool pieces and required tools used in station procedures were complete and in their specified locations. The inspectors also verified that infrequently operated flood mode pumps were in good working order, that maintenance and testing was current, and that minor deficiencies were identified in the licensee's corrective action program (CAP) with scheduled completion dates. This review constituted one inspection sample. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2. Summer Readiness of Offsite and Alternate AC Power Systems

a. Inspection Scope

In early July 2014, the inspectors performed the annual review of the licensee's readiness of offsite and alternate AC power systems prior to the onset of the high grid loading season. The inspectors reviewed procedures affecting these areas and the communications protocols between the transmission system operator and the licensee to verify that appropriate information is exchanged when issues arise that could impact the offsite power system. The inspectors walked down offsite power supply systems and emergency diesel generators, reviewed CAP documents, and interviewed appropriate plant personnel to assess deficiencies and plant readiness for summer high grid loading.

This walk down included a detailed visual inspection of the 500 kilovolt (KV) and 161 KV switchyards. Documents reviewed are listed in the Attachment. The inspectors completed one sample.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial System Walkdown

a. Inspection Scope

The inspectors performed partial walkdowns of the following four systems to verify the operability of redundant or diverse trains and components when safety equipment was inoperable. The inspectors focused on identification of discrepancies that could impact the function of the system and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, walked down control system components, and determined whether selected breakers, valves, and support equipment were in the correct position to support system operation. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the CAP. The inspectors completed four samples.

  • 2A charging system (CHS) while the 2B CHS pump was OOS

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Fire Protection Tours

a. Inspection Scope

The inspectors conducted a tour of the five areas important to safety listed below to assess the material condition and operational status of fire protection features. The inspectors evaluated whether: combustibles and ignition sources were controlled in accordance with the licensee's administrative procedures; fire detection and suppression equipment was available for use; passive fire barriers were maintained in good material condition; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with the licensee's fire plan.

Documents reviewed are listed in the Attachment. The inspectors completed five samples.

  • Vital Battery Rooms I, II, III, IV
  • Auxiliary Building elevation 669 (U1 & U2 CCP, SI)
  • ERCW Pumping Station
  • Auxiliary Building elevation 714 (General Areas)

b. Findings

No findings were identified.

.2 Annual Drill Observations

a. Inspection Scope

On September 16, the inspectors observed an unannounced fire drill at the hydrogen trailer. The inspectors assessed fire alarm effectiveness; response time for notifying and assembling the fire brigade; the selection, placement, and use of firefighting equipment; use of personnel fire protective clothing and equipment (e.g., turnout gear, self-contained breathing apparatus); communications; incident command and control; teamwork; and firefighting strategies. The inspectors also attended the post-drill critique to assess the licensee's ability to review fire brigade performance and identify areas for improvement. Following the critique, the inspectors compared their findings with the licensee's observations and to the requirements specified in the licensee's Fire Protection report. This activity constituted one inspection sample.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

.1 Internal Flooding

a. Inspection Scope

The inspectors reviewed related flood analysis documents and walked down the area listed below containing risk-significant structures, systems, and components susceptible to flooding. The inspectors verified that plant design features and plant procedures for flood mitigation were consistent with design requirements and internal flooding analysis assumptions. The inspectors also assessed the condition of flood protection barriers and drain systems. In addition, the inspectors verified the licensee was identifying and properly addressing issues using the corrective action program. The inspectors completed one sample. Documents reviewed are listed in the Attachment.

  • 1A EDG Room and Diesel Building Corridor

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

.1 Quarterly Review

a. Inspection Scope

The inspectors performed one licensed operator requalification program review. The inspectors observed a simulator session on September 4, 2014. The training scenario involved a Seal Water Return Heat Exchanger Leak with a Faulted Steam Generator. The inspectors observed crew performance in terms of: communications; ability to take timely and proper actions; prioritizing, interpreting and verifying alarms; correct use and implementation of procedures, including the alarm response procedures; timely control board operation and manipulation, including high risk operator actions; oversight and direction provided by shift manager, including the ability to identify and implement appropriate Technical Specification (TS) action; and, group dynamics involved in crew performance. The inspectors also observed the evaluators' critique and reviewed simulator fidelity to verify that it matched actual plant response. Documents reviewed are listed in the Attachment. This activity constituted one inspection sample.

b. Findings

No findings were identified

.2 Quarterly Review of Licensed Operator Performance

a. Inspection Scope

The inspectors observed and assessed licensed operator performance in the main control room during periods of heightened activity or risk. The inspectors reviewed various licensee policies and procedures such as OPDP-1, Conduct of Operations, NPG-SPP-10.0, Plant Operations, and 0-GO-5, Normal Power Operation. The inspectors utilized activities such as post-maintenance testing, surveillance testing, unplanned transients, infrequent plant evolutions, plant startups and shutdowns, reactor power and turbine load changes, and refueling and other outage activities to focus on the following conduct of operations as appropriate:

  • operator compliance and use of procedures
  • control board manipulations
  • communication between crew members
  • use and interpretation of plant instruments, indications and alarms
  • use of human error prevention techniques
  • documentation of activities, including initials and sign-offs in procedures
  • supervision of activities, including risk and reactivity management
  • pre-job briefs Specifically, the inspectors observed licensed operator performance during the following activities:
  • reactivity manipulation involving the addition of boric acid in the RCS Documents reviewed are listed in the Attachment. This activity constituted one inspection sample.

b. Findings

No findings were identified

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the maintenance issue listed below to verify the effectiveness of the licensee's activities in terms of: appropriate work practices; identifying and addressing common cause failures; scoping in accordance with 10 CFR 50.65(b);

characterizing reliability issues for performance; trending key parameters for condition monitoring; charging unavailability for performance; classification in accordance with 10 CFR 50.65(a)(1) or (a)(2); appropriateness of performance criteria for structure, system, or components (SSCs) and functions classified as (a)(2); and appropriateness of goals and corrective actions for SSCs and functions classified as (a)(1). Documents reviewed are listed in the Attachment. The inspectors completed one sample.

  • CDE #2756 - Failure of Containment Vacuum breaker 2-VLV-30-573

b. Findings

Introduction.

A Self-revealing Green NCV of Technical Specification 6.8.1 was noted for the licensee's failure to adequately implement a maintenance procedure associated with a vacuum relief containment isolation valve. Specifically, during a refueling outage on May 24, 2014, the licensee failed to properly install a locking wire associated with the spring tension bolts on the Unit 2 containment vacuum relief valve. This error ultimately led to a failure of the valve on June 24 at 1600 and entry into TS 3.6.3, "Containment Isolation Valves." The valve was ultimately repaired and the valve was declared operable on June 26 at 0026.

Description.

On June 24 at 1600, the main control began a routine purge of the containment atmosphere. The purge was secured at 1647. At 2125, Operations personnel noted the vacuum relief valve, 2-VLV-30-573 was not in its normally closed position. This was evidenced by two of the three limit switches indicated the valve was partially open. Following a review of the issue, operations declared the containment isolation function of the vacuum breaker to be inoperable at 2315 and entered TS LCO 3.6.3 action 'a' which required the penetration to be isolated in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Maintenance personnel were dispatched to the valve and noted that indeed the valve was cracked open. Maintenance personnel adjusted the valve and established adequate seating and action 'a' was exited at 0311 on June 25. However, the vacuum breaker was deemed inoperable and thus LCO 3.6.6 for the vacuum breaker was entered. This LCO had an action time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. VLV-30-573 was subsequently repaired, retested and declared back in service at 0026 on June 26.

An apparent cause evaluation revealed that a locking device (a lock wire) was not appropriately installed on the spring tension bolts inside the valve on May 24 during the last Unit 2 refueling outage. Instead, the lock wire was misapplied to the valve body. This ultimately led to a slow degradation of the valve as the spring nut was allowed to gradually back out and thus reduce the closing spring pressure on the valve. During repair of the valve, it was noted that the as-found spring pressure on the valve was approximately 17 psi while the as-left was nominally 30 psi. The perturbation of the

containment purge activity lowered containment pressure enough to momentarily actuate the valve open and the reduced closing spring force allowed the valve to become misaligned as it reclosed. At this point, the valve was partially open and considered failed as it could no longer perform its containment isolation function of closing.

Analysis.

The licensee's failure to adequately develop and implement a procedure governing the maintenance of a containment isolation valve was a performance deficiency. Specifically, failure to follow a work order package precluded the proper installation of a lock wire inside a containment isolation valve. This then led to the gradual loosening of the spring nut which led to a reduction of the closing spring force on the valve. The containment purge evolution led to the ultimate failure of the valve during the momentary opening and closing of the valve. The inspectors evaluated this issue in accordance with the NRC's significance determination process (SDP). This finding was determined to be greater than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstone's objective to ensure the structural integrity of the containment boundary. Specifically, the finding challenged containment integrity. Using IMC 0609.04, Initial Characterization of Findings and IMC 0609 Appendix A, Exhibit 3 - Barrier Screening Questions, the finding required analysis using MC 0609, Appendix 'H', "Containment Integrity Significance Determination Process," as the finding represented an open pathway in the physical integrity of containment via a valve. According to Appendix H, the finding represented a Type B finding as it did not directly affect the core damage frequency. The finding was then processed under Section 6 using a phase 2 analysis. The inspectors performed the phase 2 analysis using Sec 6.2 of Appendix H. The inspectors determined, based on the actual circumstances being evaluated, that a more detailed assessment needed to be performed in a SDP Phase 3 evaluation. The regional Senior Reactor Analyst performed a Phase 3 SDP for the finding. A screening analysis was conducted using the assumption that all core damage sequences would lead to a Large Early Release. This was an overestimation of risk, since actions to mitigate a release were possible.

The short exposure time multiplied by the Core Damage Frequency for the plant resulted in less than a 1E-7 increase in Large Early Release Probability, and the finding is GREEN. The cause of this finding was determined to have a cross-cutting aspect in the Human Performance component, relating to the assurance by supervision that procedures are adequate to ensure nuclear safety. [H.1].

Enforcement:

Unit 2 TS 6.8.1.a. required, in part, that written procedures be established, implemented, and maintained covering the activities specified in Appendix A, "Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors," of Regulatory Guide (RG) 1.33, "Quality Assurance Program Requirements (Operations)," Revision 2, dated February 1978. RG 1.33 Appendix A Section 9.a, "Procedures for Performing Maintenance," required, in part, that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. The work instructions from WO 07-779292-000 (step 5.7.21) required installation of the safety wire (lock wire) but did not provide any detail for how this was to be accomplished. Contrary to the above, on May 24, 2014, the licensee failed to properly install the safety wire such that the spring nut was precluded from loosening. Specifically, the maintenance procedure failed to provide adequate detail and guidance to ensure the safety wire was properly installed to ensure the spring nut was adequately locked. Corrective actions included: 1) repair of the vacuum relief valve; 2) the development of a maintenance case study for inclusion into maintenance training; and 3) Revision of the operating instruction to provide validation of containment vacuum relief valve position following containment purges. Because the finding was of very low safety significance and has been entered into the licensee's CAP as PER 902721, this violation is being treated as an NCV, consiste nt with the NRC Enforcement Policy: NCV 05000328/2014004-01, Failure of Containment Vacuum Relief.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the following activities to determine whether appropriate risk assessments were performed prior to removing equipment from service for maintenance. The inspectors evaluated w hether risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent work was performed, the inspectors reviewed whether plant risk was promptly reassessed and managed. The inspectors also assessed whether the licensee's risk assessment tool use and risk categories were in accordance with Standard Programs and Processes Procedure NPG-SPP-07.1, "On-Line Work Management," Revision 3, and Instruction 0-TI-DSM-000-007.1, "Risk Assessment Guidelines," Revision 9. The inspectors completed five samples.

  • emergent work 2A EDG Lube oil recirculation pump replacement
  • emergent failure of overcurrent 51 relay on 1B CHS pump breaker
  • yellow Risk due to 1A main control room (MCR) chiller work

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

For the operability evaluations described in the problem event reports (PERs) listed below, the inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and t he subject component or system remained available, such that no unrecognized increase in risk occurred. The inspectors compared the operability evaluations to the Updated Final Safety Analysis Report (UFSAR) descriptions to determine if the system or component's intended function(s) were adversely impacted. In addition, the inspectors reviewed compensatory measures implemented to determine whether the compensatory measures worked as stated and the measures were adequately controlled. The inspectors also reviewed a sampling of PERs to assess whether the licensee was identifying and correcting any deficiencies associated with operability evaluations. The inspectors completed four samples.

  • PER 907134, prompt determination of operability (PDO) on 2A Pipe Chase Cooler leak
  • PER 922622, PDO on Unit 1 Ice Condenser Icing
  • PER 914789, Foreign material (10" rubber material) found in main turbine oil strainer
  • SRs 932083 & 932087 associated with U2 Eagle 21 TSP failure

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Temporary Modifications

a. Inspection Scope

The inspectors reviewed the temporary modification listed below and the associated 10 CFR 50.59 screening, and compared it against the UFSAR and TS to verify whether the modification affected operability or availability of the affected system.

  • TACF - SQN-1-2014-085-001 - DC Wetting Current Source Installation for Rod Position Indicator (RPI) E-5 Following installation and testing, the inspectors observed indications affected by the modification, discussed them with operators, and verified that the modification was installed properly and its operation did not adversely affect safety system functions.

The inspectors completed one sample.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing (PMT)

a. Inspection Scope

The inspectors reviewed the post-maintenance tests associated with the work orders (WOs) listed below to assess whether procedures and test activities ensured system operability and functional capability. The inspectors reviewed the licensee's test procedure to evaluate whether: the procedure adequately tested the safety function(s) that may have been affected by the maintenance activity; the acceptance criteria in the procedure were consistent with information in the applicable licensing basis and/or design basis documents; and the procedure had been properly reviewed and approved. The inspectors also witnessed the test or reviewed the test data to determine whether test results adequately demonstrated restoration of the affected safety function(s). The inspectors completed five samples.

  • 116008676, Restore from Abnormal Operating Procedure and Verify Operability of Drawer N-44A
  • 116027085, Replace 50/51 OC relays on A & C phase of 1B CHS Pump
  • 115603969, Calibrate Auxiliary Air Compressor A-A oil level switch
  • 116156226, U2 Stroking valves for PMT or release from HO
  • 115725379, RHR HX Outlet valve

b. Findings

No findings were identified.

1R20 Refueling and Outage Activities

a. Inspection Scope

Unit 2 Refueling Outage Heat-up and Start-up Activities

To verify reactor coolant safety (RCS) integrity and containment integrity, the inspectors further reviewed the licensee's RCS leakage calculations and containment isolation valve lineups. In order to verify that core operating limit parameters were consistent with core design, the inspectors also observed portions of the low power physics testing, including reactor criticality. The inspectors also reviewed the physics testing results. The inspectors completed one sample.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

For the surveillance tests identified below, the inspectors assessed whether the structures, systems, and components (SSCs) involved in these tests satisfied the requirements described in the TS surveillance requirements, the UFSAR, applicable licensee procedures, and whether the tests demonstrated that the SSCs were capable of performing their intended safety functions. This was accomplished by witnessing testing and/or reviewing the test data. The inspectors completed three samples.

Routine Surveillance Tests

  • 2-SI-EDC-082-307.A, Undervoltage/Degraded Voltage EDG Start and Load Shedding Time Response Relay Test, Revision 17
  • 1-SI-OPS-082-024.A, 1A-A DG 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Run and Load Rejection Test, Rev. 26 RCS leakage test

b. Findings

No findings were identified.

Cornerstone:

Emergency Preparedness 1EP6 Drill Evaluation

a. Inspection Scope

Resident inspectors evaluated the conduct of routine licensee emergency drill on July 8, and August 13, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation (PAR) development activities. The inspectors observed emergency response operations in the simulated control room to verify that event classification and notifications were done in accordance with EPIP-1, Emergency Plan Classification Matrix, Revision 51. The inspectors also attended the licensee critique of the drill to compare any inspector observed weakness with those identified by the licensee in order to verify whether the licensee was properly identifying deficiencies.

The inspectors completed two samples.

b. Findings

No findings were identified.

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security

OTHER ACTIVITIES

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

The inspectors sampled licensee submittals for the Performance Indicators (PI) listed below for the period from July 1, 2013, through June 30, 2014 for both Unit 1 and Unit 2. Definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Indicator Guideline, Revision 6, were used to determine the reporting basis for each data element in order to verify the accuracy of the PI data reported during that

period.

Cornerstone:

Mitigating Systems

  • Safety System Functional Failures The inspectors reviewed portions of the operations logs and raw performance index (PI)data developed from monthly operating reports and discussed the methods for compiling and reporting the PIs with engineering personnel. The inspectors also independently calculated selected reported values to verify their accuracy and compared graphical representations from the most recent PI report to the raw data to verify that the data was correctly reflected in the report. Specifically for the Mitigating Systems Performance Index (MSPI), the inspectors reviewed the basis document and derivation reports to verify that the licensee was properly entering the raw data as suggested by NEI 99-02.

For Safety System Functional Failures, the inspectors also reviewed licensee event reports (LERs) issued during the referenced timeframe. The inspectors completed 12 samples (six per unit).

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

.1 Daily Review

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensee's CAP. This was accomplished by reviewing the description of each new PER and attending daily management review committee meetings.

b. Findings and Observations

No findings were identified.

.2 Selected Issue Follow-up:

Four NCVs with Cross Cutting Aspect of Evaluation

a. Inspection Scope

The inspectors performed an in-depth review of PER 876781. In April 2014, the licensee noted there were four NRC-issued NCVs received within the previous 12 months that exhibited the cross cutting aspect of "eval uation" as described in NRC manual chapter (MC) 0310, "Aspects Within Cross Cutting Areas." The cross cutting aspects were part of four separate issues (NCVs): 1) Failure to evaluate motor operated valves with regard to lower degraded voltage values, 2) Failure to correct in a timely manner a dual indication on residual heat removal (RHR) valve, 1-FCV-63-72, 3) Failure to install sprinklers in accordance with the fire protection program (FPP) and code requirements, and 4) Failure to perform the required review when adding fire watches to the FPP. MC 0310 describes a cross cutting aspect as a performance characteristic of a finding that is the most significant causal factor of the performance deficiency. Specifically, the four findings above were related to "evaluations" (Section 06.02.P.2 under MC 0310) and noted to contain licensee weaknesses regarding the thorough evaluation of issues to ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance.

As a result of this identified weaknes s, the licensee developed an Apparent Cause Evaluation (ACE). The inspectors reviewed the actions taken to determine if the licensee had adequately addressed the following attributes:

  • complete, accurate and timely identification of the problem
  • evaluation and disposition of operability and reportability issues
  • consideration of previous failures, extent of condition, generic or common cause implications
  • prioritization and resolution of the issue commensurate with safety significance
  • identification of the root cause and contributing causes of the problem
  • identification and implementation of corrective actions commensurate with the safety significance of the issue

b. Findings and Observations

The inspectors noted the ACE to be thorough and probing. The conclusion of the ACE was that the staff had knowledge deficiencies in 1) documenting critical thinking; 2)understanding design and licensing bases; and 3) identifying potential impacts to nuclear

safety and compliance when performing required problem evaluations. The corrective actions included training and stand-downs to stress the importance of thoroughly evaluating significant safety and regulatory issues. The inspectors also noted that the licensee performed a detailed extent of condition as well as an extent of cause analysis

as part of this ACE.

.3 Selected Issue Follow-up:

Operator Requalification Training Corrective Actions

a. Inspection Scope

The inspectors performed an in-depth review of PER 909616 associated with the licensed operator training program. In July 2014, during performance of licensed operator requalification training there were issues identified with the critical task standards in the approved simulator scenario. The inspectors reviewed the actions taken to determine if the licensee had adequately addressed the following attributes.

  • complete, accurate and timely identification of the problem
  • evaluation and disposition of operability and reportability issues
  • consideration of previous failures, extent of condition, generic or common cause implications
  • prioritization and resolution of the issue commensurate with safety significance
  • identification of the root cause and contributing causes of the problem
  • identification and implementation of corrective actions commensurate with the safety significance of the issue

b. Findings and Observations

No findings were identified.

4OA3 Event Follow-up

.1 (Closed) Licensee Event Report (L

ER) 05000328/2014-002-00, Containment Vacuum Relief Valve Inoperable Resulting in Condition Prohibited by Technical Specifications

a. Inspection Scope

On June 24, 2014, at 2125 a MCR operator noted that the containment vacuum relief valve 2-VLV-30-573 was not in its normally closed position. The initial characterization of this deficiency was that the position indication was faulty but the valve was still operable. This caused the crew to enter TS LCO 3.6.3 action 'c' (Similarly, LCO 3.6.6). Both LCOs were 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and provided the crew time to deal with the issue. However, based on further review of the issue and telephone conversations with management, the crew concluded that, indeed, the valve was inoperable at 2315. This was based on the fact that a containment purge had been performed earlier in the day and was concluded at 1647. The licensee concluded that based on past experience, a vacuum relief can partially open due to localized low pressure areas that may result when purging the containment. However, it was surmised that in this instance, the vacuum relief failed to properly reseat after momentarily opening as evidenced by one of the three closing limit switched failing to actuate. This defeated the containment isolation function of the penetration itself and thus required entry into LCO 3.6.3 action 'a' which is a four hour LCO to isolate the penetration. The crew ultimately isolated the penetration at 0247 on June 25 by jacking closed and de-powering the vacuum relief isolation valve. The vacuum relief was ultimately repaired and placed back in to service at 0256 on June 26.

Further review of this event by the licensee, noted that the conditions were readily available at 2125 that the vacuum relie f was partially open and the containment penetration was inoperable. Additionally, the licensee concluded that most likely the vacuum relief became inoperable at 1600 when the purge was started. However, the penetration was not isolated until 0311 the following day for a total of slightly more than 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> of inoperability. Considering the LCO action time of four hours to isolate the penetration along with the six hours action time, the total time to be in mode 3 for LCO 3.6.3 action 'a' is 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Thus, in this instance, failing to isolate the penetration in four hours or placing the Unit 2 in mode 3 in the following six hours (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> total) was a condition prohibited by TS.

Thus, this LER was required pursuant to 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by plant technical specifications.

b. Findings

NCV 05000328/2014004-01 was issued as a result of this performance deficiency and is documented in this report under Section 1R12.

.2 (Closed) Licensee Event Report (LER) 05000327/2014-001-00, Never Performed Technical Specification Surveillance for Common Spare Component Cooling System (C-S CCS) Pump

a. Inspection Scope

On December 19, 2013, an operator trainee noted a lack of procedural guidance to perform TS Surveillance Requirement (SR) 4.7.3.b which required the testing of the

safety injection system (SIS) start signal to the C-S CCS pump every 18 months. This pump is normally powered from the 480 volt 'A' train safety-related bus on Unit 2 and alternatively can be powered from the 480 volt 'B' train safety-related bus on Unit 1.

Based on design documents, the SIS signal from either unit should give a start signal to the C-S CCS pump. Since the pump is normally powered from Unit 2, the SIS start signal from Unit 2 was tested every 18 months per procedure 2-SI-OPS-082-026.B, "Loss of Offsite Power with Safety Injection - D/G 2B-B Test." This issue was entered into the CAP as PER 826482.

An initial investigation into this issue determined the lack of testing to be a missed surveillance. Operations invoked TS SR 4.0.3 as the total functionally required by SR 4.7.3.b was not fully tested. Since the surveillance frequency of SR4.7.3.b was more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (18 months) and the expected retest was to be in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time of discovery, the licensee performed a risk evaluation on how to manage the increased risk of an untested component. On January 13, 2014, the performance of SIS start testing from both units to the C-S CCS pump was completed and SR 4.7.3.b was

then met.

Subsequent investigation completed on January 30, revealed that, in fact, the C-S pump was never tested in the Unit 1 configuration and thus the deficiency was reclassified as a never performed surveillance. Thus, this LER was required pursuant to 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by plant technical specifications as SR 4.7.3.b was deemed never to have been performed.

The inspectors discussed the event with operations, maintenance, engineering, and licensee management personnel to gain an understanding of the conditions leading up to the event and assess licensee actions taken following the event. Additionally, the inspectors reviewed the apparent cause evaluation report to assess the detail and thoroughness of the evaluation and the adequacy of the proposed corrective actions.

b. Findings

This LER resulted in a licensee identified violation which is described in Section 4OA7.

.3 (Closed) Licensee Event Report (LER) 05000327/2014-002-00, Lack of Administrative Controls for Some Containment Penetrations During Fuel Movement Results in Condition Prohibited by Technical Specifications

a. Inspection Scope

On June 6, 2014, the licensee completed a review of administrative control of penetrations used for ice condenser maintenance during fuel movement. The licensee concluded that they had been inappropriately crediting engineering evaluations as justification to utilize open penetrations during fuel movement. In 2000, the licensee obtained TS change 99-15 to TS 3.9.4 that permitted some containment penetrations to remain open during fuel movement. Prior to 2000, the licensee used engineering safety evaluation to justify open penetrations during fuel movement. However, in 1999, the NRC deemed the practice of using 10 CFR 50.59 (via engineering evaluation) as inappropriately affected TS 3.9.4 without prior NRC approval and the practice was ceased. Due to a failure to correct a plant procedure 1,2-SI-OPS-088-006.0, "Containment Building Ventilation Isolation," the practice was inadvertently continued from 2000 to 2014 with respect to four containment penetrations: 1) X79A (transfer ice to ice condenser); X79B (transfer ice to ice condenser); X54 (steam generator sludge lancing) and X117 (ice condenser condensate drain). This LER was submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by technical specification.

The inspectors discussed the event with operations, maintenance, engineering, and licensee management personnel to gain an understanding of the conditions leading to the event and assess licensee actions taken following the event. Additionally, the inspectors reviewed the apparent cause evaluation report to assess the detail and thoroughness of the evaluation and the adequacy of the proposed corrective actions.

One of the corrective actions was to upgrade plant procedures to ensure strict control of penetration during fuel movement and the elimination of the engineering evaluation avenue to comply with TS 3.9.4.

The inspectors reviewed the LER and PER 886970 to verify that the cause of the breeched penetration was identified and whether corrective actions were appropriate.

The licensee's ACE identified that the controlling procedure did not reference that the four affected penetrations were under administrative control during fuel movement. However, there were other procedures in place that in all likelihood would have caused personnel to isolate these penetrations given a fuel handling accident. Notwithstanding the above, strict TS 3.9.4 compliance was not met regarding the above penetrations.

This LER was submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by technical specification.

b. Findings

This LER resulted in a licensee identified violation which is described in Section 4OA7.

4OA5 Other Activities

a. Inspection Scope

(Closed) VIO 05000327,328/2012005-05, Failure to Adequately Evaluate and Qualify Molded Case Circuit Breakers The inspectors reviewed various documents associated with the above violation. This violation dealt with the seismic mounting characteristics of Heinemann circuit breakers which are used in 120volt vital AC applications. The violation discussed several discrepancies between the physical mounting of these circuit breakers in safety related cabinets and the design documents. The violation also questioned the seismic adequacy of the current mounting design. The licensee analyzed the issue in PER 668367 and formalized a consistent mounting design using Micarta board to ensure adequate seating force for the circuit breakers. The licensee tested the new configuration in a laboratory environment to verify the design could withstand an operating basis earthquake as well as a safe shutdown earthquake. The licensee then performed a plant modification, DCN 23189, "Improvements to Design and Documentation for 120 VAC Vital Instrument Power Board Heinemann Breaker Mountings," Rev A. This modification covered 48 circuit breakers on both units and was completed on Unit 1 during the fall 2013 and Unit 2 during the spring 2014. The inspectors reviewed this modification as well as the associated work orders.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

.1 Exit Meeting Summary

On October 2, 2014, the resident inspectors presented the inspection results to Mr. Carlin and other members of his staff, who acknowledged the findings. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.

4OA7 Licensee-identified Violations

The following violations of very low significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as NCVs.

.1 Failure to perform a TS required surveillance

Unit 1 SR 4.7.3.b requires in part that at least once per 18 months, each CCS pump start automatically on a SIS signal. Contrary to the above, the C-S (swing) CCS pump was never tested to automatically start using a Unit 1 SIS signal. The licensee considered this to be a never performed surveillance rather than simply a missed surveillance since no testing records since initial operation in 1980 could be located. The pump had been successfully tested every 18 months from the Unit 2 SIS as this pump is normally powered from Unit 2. As an alternate and abnormal lineup, the pump could be powered from Unit 1 and thus the requirement for start testing using the Unit 1 SIS. This finding was considered more than minor because it was associated with the mitigating system cornerstone and affected the cornerstone's reliability due to the failure to fully test the C-S CCS pump for over 30 years. The finding was considered of very low safety significance as it remained operable and available during the affected period due to successfully passing the initial surveillance test performed on January 13, 2014. The issue was entered into the CAP as PER 826482. Note that this issue is also discussed under Section

4OA3 of this report as it involved a LER.

.2 Lack of Administrative Controls Over Containment Penetrations During Fuel Movement

TS 3.9.4.c requires in part that during fuel movement, all containment penetrations shall be closed or capable of being closed by an automatic valve. In addition, TS 3.9.4.c allows exceptions to this requirement for penetrations that traverse to the auxiliary building secondary containment enclosure (ABSCE) where these penetrations may be open under administrative controls during fuel movement. Contrary to the above, on several instances between 2000 and 2014, the licensee opened penetrations (between containment and ABSCE) during fuel movement without adequate administrative controls in place. The finding was considered more than minor because it was associated with the Barrier Integrity cornerstone and affected the cornerstone's ability to preserve the containment boundary. The inspectors determined that, although the finding involved a violation of the containment control, TS 3.9.4, the finding did not:

1) involve a loss of reactor coolant system (RCS) inventory; 2) degrade ability to terminate a leak path or add RCS inventory as needed; or 3) degrade the ability to recover RHR once it was lost. Therefore, according Appendix G, the finding did not require a quantitative (phase 2 or 3) analysis. Findings in the shut-down condition that do not require a quantitative analysis are considered to be of very low safety significance (Green). This issue was entered into the CAP as PER 886970. Note that this issue is also discussed under Section

4OA3 of this report as it involved a LER.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

J. Carlin, Site Vice President
A. Day, Chemistry Senior Manager
D. Erb, Work Management Director
E. Henderson, Licensing Manager
J. Johnson, Program Manager Licensing
A. Little, Site Security Manager
T. Marshall, Operations Director
P. Noe, Plant Support Director
W. Pierce, Site Engineering Director
P. Pratt, Plant Maintenance Manager
M. Purcell, Quality Assurance Manager
P. Simmons, Plant Manager

NRC personnel

A. Hon, Project Manager, Office of Nuclear Reactor Regulation

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000328/2014004-01 NCV Failure to Perform Adequate Maintenance

on Containment Vacuum Relief Valve

(Section 1R12)

Closed

05000327,328/2012005-05 VIO Failure to Adequately Evaluate and Qualify
Molded Case Circuit Breakers (Section
4OA5)
05000327/LER-2014-001-00 LER Never Performed TS Surveillance for C-S
CCS Pump (Section 4OA3)
05000328/LER-2014-002-00 LER Containment Vacuum Relief Valve Inoperable Resulting in Condition Prohibited

by TS (Section 4OA3)

05000327/LER-2014-002-00 LER Lack of Administrative Controls for Some Containment Penetrations During Fuel Movement Results in Condition Prohibited

by Technical Specifications (Section 4OA3)

LIST OF DOCUMENTS REVIEWED

Section R01:

Adverse Weather Protection

Procedures

AOP-N.03, Flooding, Revision 32
0-PI-OPS-510-001.0, Flood Preparation Equipment Inventory, Revision 8
Section R05:
Fire Protection

Procedures

FPDP-1, Conduct of Fire Protection, Revision 2
0-PI-FPU-317-299.W, Att. 8, Shift Check List, Revision 32
NPG-SPP-18.4.7, Control of Transient Combustibles, Rev. 0
EITP-100, Environmental Compliance, Rev. 6 0-SI-FPU-410-703.0, Inspection of FPR Required Fire Doors, Rev. 5
SQN-FPR-Part-II, SQN Fire Protection Report Part II - Fire Protection Plan, Revision 28
Section R06:
Flood Protection Measures

Work Orders

WO 11108121224, Check Standing Water Level in Manholes/Handholes
Other documents
TVA letter to NRC dated May 4, 2007. TVA response to
GL 2007-01
Section R11:
Licensed Operator Requalification

Procedures

NPG-SPP-17.8.1 rev 10, 11 Licensed Operator Requalification Examination Development and Implementation
NPG-SPP-22.207 rev 0 Procedure Use and Adherence
PERs
PER 907424
PER 909616
Other documents
14-3 CPE Crew documentation and remediations OPL273E1403 R1 & 2
WCAP-17711-NP1 Pressurized Water Reactor Owners Group Westinghouse Emergency Response Guideline Revision 2-Based Critical Tasks
Section R12:
Maintenance Effectiveness

Procedures

TI-4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -
10CFR50.65, Revision 23
Other documents
CDE #2756 - Failure of Containment Vacuum breaker 2-VLV-30-573
Attachment Section R13:
Maintenance Risk Assessments and Emergent Work Evaluation

Procedures

0-TI-DSM-000-007.1, Risk Assessment Guidelines, Revision 9
NPG-SPP-07.3, Work Activity Risk Management Process, Revision 3
NPG-SPP-07.2.4, Forced Outage or Short Duration Planned Outage Management, Revision 0
NPG-SPP-07.2, Outage Management, Revision 0
GOI-6, Apparatus Operations, Revision 142
Section R15:
Operability Evaluations

Procedures

NEDP-22, Functional Evaluations, Rev. 9
OPDP-8, Limiting Conditions for Operation Tracking, Rev. 5
NPG-SPP-03.5, Regulatory Reporting Requirements, Revision 2
PER/SRs
PER 907134, prompt determination of operability (PDO) on 2A Pipe Chase Cooler leak
PER 922622, PDO on Unit 1 Ice Condenser Icing
PER 914789, Foreign material (10" rubber material) found in main turbine oil strainer SRs
932083 &
932087 associated with U2 Eagle 21 TSP failure Section R18:
Plant Modifications

Procedures

NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Revision 4
NPG-SPP-09.4, 10
CFR 50.59 Evaluations of Changes, Tests, and Experiments, Revision 1
NPG-SPP-09.5, Temporary Alterations, Revision 0
Other TACF -
SQN-1-2014-085-001 - DC Wetting Current Source Installation for Rod Position Indicator (RPI) E-5
Section R19:
Post Maintenance Testing

Procedures

MMDP-1, Maintenance Management System, Revision 20
MMDP-3, Guidelines for Planning and Execution of Troubleshooting Activities, Revision 6
NPG-SPP-6.5, Foreign Material Control, Revision 0
NPG-SPP-6.1, Work Order Process Initiation, Revision 0
NPG-SPP-06.3, Pre-/Post-Maintenance Testing, Revision 0
NPG-SPP-06.9, Testing Programs, Revision 0
NPG-SPP-06.9.1, Conduct of Testing, Revision 1
NPG-SPP-06.9.3, Post-Modification Testing, Revision 0

Work Orders

116008676, Restore from Abnormal Operating Procedure and Verify Operability of Drawer
N-44A
116027085, Replace 50/51 OC relays on A & C phase of 1B CHS Pump
115603969, Calibrate Auxiliary Air Compressor A-A oil level switch
116156226, U2 Stroking valves for PMT or release from HO
115725379, RHR HX Outlet valve Attachment Section R20:
Refueling and Outage Activities

Procedures

FHI-3, Movement of Fuel, Revision 65
0-GO-15, Containment Closure Control, Revision 34
0-GO-13, Reactor Coolant System Drain and Fill Operations, Revision 71
NPG-SPP-08.1, Nuclear Fuel Management, Revision 00
0-PI-OPS-000-011.0, "Containment Access Control During Modes 1-4, Revision 1
Section R22:
Surveillance Testing

Procedures

2-SI-EDC-082-307.A, Undervoltage/Degraded Voltage EDG Start and Load Shedding Time
Response Relay Test, Revision 17 1-SI-OPS-082-024.A, 1A-A
DG 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Run and Load Rejection Test, Revision 26 0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory, Revision 34

Section 4OA1: Performance Indicator Verification

Procedures

NPG-SPP-02.2, Performance Indicator Program, Revision 2
NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6

Section 4OA5: Other Activities

DCN 23189, "Improvements to Design and Documentation for 120 VAC Vital Instrument Power Board Heinemann Breaker Mountings," Rev A

LIST OF ACRONYMS

ACE apparent cause evaluation
ABSCE auxiliary building secondary containment enclosure
ACR auxiliary control room
ADAMS Agencywide Documents Access and Management System
AFW auxiliary feedwater
AOP abnormal operating procedures
CAP corrective action program
CAQ condition adverse to quality
CFR Code of Federal Regulations
CHS charging system
CS containment spray
EDG emergency diesel generator
EPIP emergency plan implementing procedures
ERCW essential raw cooling water
FPP fire protection program
IMC inspection manual chapter
IP inspection procedure
KV kilovolt
LCO limiting condition for operation
LER licensee event report
NCV non-cited violation
NEI Nuclear Energy Institute
NRC U.S. Nuclear Regulatory Commission
OOS out-of-service
PARS Publicly Available Records
PER problem evaluation report
PI performance indicator Rev. Revision
RCS reactor coolant system
RG regulatory guide
RHR residual heat removal
RPI rod position indicator
RTP rated thermal power
SDP significance determination process
SER safety evaluation report
SIS safety injection system
SSC structure, system, or component
TI temporary instruction

TS technical specification TVA Tennessee Valley Authority

U1 Unit 1

U2 Unit 2

UFSAR updated final safety analysis report
WO work order
Retrieved from "https://kanterella.com/w/index.php?title=IR_05000327/2014004&oldid=786255"