IR 05000348/2013005

IR 05000348-13-005, 05000364-13-005, 05000348-13-502, and 05000364-13-502, on October 1, 2013, Through December 31, 2013; Joseph M. Farley Nuclear Plant, Units 1 and 2, Problem Identification and Resolution

ML14045A340
Person / Time
Site:	Farley
Issue date:	02/14/2014
From:	Croteau R Division Reactor Projects II
To:	Gayheart C Southern Nuclear Operating Co
References
EA-12-145, EA-14-017 IR-13-005
Download: ML14045A340 (40)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION ary 14, 2014

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000348/2013005; 05000364/2013005; 05000348/2013502; AND 05000364/2013502, AND NOTICE OF VIOLATION

Dear Ms. Gayheart:

On December 31, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Joseph M. Farley Nuclear Plant, Units 1 and 2. On January 30, 2014, the NRC inspectors discussed the results of this inspection with you and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

The enclosed inspection report discusses a finding of low-to-moderate safety significance (White). As described in Section 4OA2.3 of the enclosed inspection report, a calculation error resulted in the radiological initiating conditions threshold values for the RG1 (General Emergency) and RS1 (Site Area Emergency) Emergency Action Levels to be sixty times greater than the appropriate values. This finding resulted in an immediate safety concern for which appropriate immediate corrective actions were taken. The correct threshold values were provided to the appropriate operations staff decision makers which resolved the immediate safety concern. Additional corrective actions taken included performing a causal determination, processing formal changes to the stations emergency plan and associated implementing procedures, and performing extent of condition/cause reviews throughout the Southern Nuclear Operating Company fleet. Following the internal review process, the revised emergency plan and associated implementing procedure were provided to the NRC in July 2013.

In a telephone conversation on January 27, 2014, Mr. Brian Bonser, Chief, Plant Support Branch, Division of Reactor Safety, Region II, informed you of the details of the preliminary finding, the apparent violation, and advised Farley representatives that the finding appeared to satisfy the old design issue criteria contained in NRC Inspection Manual Chapter 0305, Operating Reactor Assessment Program, Section 11.05, Treatment of Items Associated with Enforcement Discretion, dated October 18, 2013. The intent of this section is to establish reactor oversight process (ROP) guidance that supports the objective of enforcement discretion, which is to encourage licensee initiatives to identify and resolve problems, especially issues that are not likely to be identified by routine efforts. Additionally, Mr. Bonser advised you that based on the above, the NRC had sufficient information, including Farleys corrective actions, to make a final significance determination and enforcement decision without the need for a Regulatory Conference or a written response from you. You indicated that Farley Nuclear Plant did not believe that a Regulatory Conference or written response was necessary.

Based on the above, the NRC has concluded that the finding is appropriately characterized as White, a finding of low to moderate safety significance. Additionally, the NRC determined that the White finding meets the criteria specified in IMC 0305 for treatment as an old design issue.

The basis for the NRCs determination included the following: (1) the issue was licensee-identified through an extent of condition review prompted by Southern Co. fleet operating experience; (2) the issue was corrected within a reasonable time after discovery; (3) the issue was not likely to be previously identified by recent ongoing licensee efforts; and (4) the issue was not reflective of a current performance deficiency associated with existing programs, policy, or procedures. Therefore, in accordance with IMC 0305, the performance issue will not aggregate in the Action Matrix with other performance indicators and inspection findings. Note IMC 0305 specifies the need for an inspection in accordance with inspection procedure (IP)

95001 Supplemental Inspection for One or Two White Inputs in a Strategic Performance Area, even if the White finding meets the criteria for treatment as an old design issue to review the licensees root cause and corrective action plans. The White finding will remain open until IP 95001 is completed.

The NRC has also determined that the failure to maintain the effectiveness of your emergency plan is a violation of 10 CFR Part 50.54(q)(2), as cited in the attached Notice of Violation (Notice). The circumstances surrounding the violation are described in detail in the enclosed inspection report. In accordance with the NRC Enforcement Policy, the Notice is considered escalated enforcement action because it is associated with a White finding.

The NRC has concluded that the information regarding the reason of the violation, the corrective actions taken to correct the violation and prevent recurrence, and the date when full compliance was achieved is already adequately addressed on the docket in the enclosed inspection report.

Therefore, you are not required to respond to this letter unless the description therein does not accurately reflect your corrective actions or your position.

No NRC-identified or self-revealing findings were identified during this inspection. However, inspectors documented two licensee-identified violations which were determined to be of very low safety significance and Severity Level IV in this report. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violations or significance of these NCVs, 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, U. S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Joseph M. Farley Nuclear Plant. As a result of the Safety Culture Common Language Initiative, the terminology and coding of cross-cutting aspects were revised beginning in calendar year (CY) 2014. New cross-cutting aspects identified in CY 2014 will be coded under the latest revision to Inspection Manual Chapter (IMC) 0310. Cross-cutting aspects identified in the last six months of 2013 using the previous terminology will be converted to the latest revision in accordance with the cross-reference in IMC 0310. The revised cross-cutting aspects will be evaluated for cross-cutting themes and potential substantive cross-cutting issues in accordance with IMC 0305 starting with the CY 2014 mid-cycle assessment review.

In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC public document room or from the publicly available records (PARS) component of NRCs 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).

Sincerely,

/RA/

Richard Croteau, Director Division of Reactor Projects Docket Nos.: 50-348, 50-364 License No.: NPF-2, NPF-8

Enclosures:

1. Inspection Report 05000348/2013005; 05000364/2013005, 05000348/2013502; AND 05000364/2013502 w/Attachment: Supplemental Information 2. Notice of Violation

REGION II==

Docket Nos.: 05000348, 05000364 License Nos.: NPF-2, NPF-8 Report No.: 05000348/2013005, 05000364/2013005, 05000348/2013502, and 05000364/2013502 Licensee: Southern Nuclear Operating Company, Inc.

Facility: Joseph M. Farley Nuclear Plant, Units 1 and 2 Location: Columbia, AL Dates: October 1, 2013 through December 31, 2013 Inspectors: P. Niebaum, Senior Resident Inspector J. Sowa, Resident Inspector M. Speck, Senior Emergency Preparedness Inspector (4OA2)

J. Laughlin, Emergency Preparedness Inspector (1EP4)

C. Dykes, Health Physicist (4OA5)

B. Pursley, Health Physicist (4OA5)

B. Caballero, Senior Operations Engineer (1R11.3)

L. Lake, Senior Reactor Inspector (1R08)

Approved by: Frank Ehrhardt, Chief Reactor Projects Branch 2 Division of Reactor Projects Enclosure 1

SUMMARY OF FINDINGS

IR 05000348/2013005, 05000364/2013005, 05000348/2013502, and 05000364/2013502;

October 1, 2013, through December 31, 2013; Joseph M. Farley Nuclear Plant, Units 1 and 2,

Problem Identification and Resolution The report covered a three-month period of inspection by resident and regional inspectors. One violation was documented in this report. The significance of inspection findings are indicated by their color (i.e. greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP) dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Components Within The Cross-Cutting Areas dated October 28, 2011. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated July 9, 2013. The NRCs program for overseeing the safe operations of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process revision 4.

Cornerstone: Emergency Preparedness

• White: A finding and associated violation of 10 CFR 50.54(q)(2) was identified by the licensee for the failure to follow and maintain the effectiveness of emergency plans which use a standard emergency classification and action level scheme.

Specifically, the licensee's emergency plan emergency action level (EAL) Category R - Abnormal Radiological RG1 (General Emergency) and RS1(Site Area Emergency) specified threshold values which were sixty times too high due to a calculation error. As immediate corrective action, the licensee provided the corrected threshold values to appropriate management and decision-makers (shift managers/emergency directors). The licensee entered this issue into the corrective action program as condition report (CR) 648187.

The performance deficiency was determined to be more than minor because it was associated with the emergency preparedness cornerstone attribute of procedure quality. It impacted the cornerstone objective because it was associated with inappropriate EAL and emergency plan changes and their adequacy to protect the health and safety of the public in the event of a radiological emergency. Specifically, the licensees ability to declare a Site Area and General Emergency would be delayed based on effluent radiation monitor values was degraded in that event classification using these radiation monitors. The finding was assessed for significance in accordance with NRC Manual Chapter 0609, Appendix B,

Emergency Preparedness Significance Determination Process, which states,

FAILURE TO COMPLY means that a program is noncompliant with a REGULATORY REQUIREMENT. The inspector determined that the situation constituted a degraded standard rather than failed risk-significant planning standard (RSPS). The issue of concern was similar to the example in Table 5.4.1 (Degraded RSPS) and was determined to be of low to moderate safety significance (White).

The violation was determined to meet the IMC 0305 criteria for enforcement discretion as an old design issue. A cross-cutting aspect was not assigned based on the elapsed time since the performance deficiency occurred and because the inspectors determined it was not reflective of current licensee performance. (Section 4OA2)

Violations of very low safety significance that were identified by the licensee have been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 started the report period in a refueling outage. On October 25, Unit 1 commenced a reactor startup and achieved 100 percent rated thermal power (RTP) on October 29. On November 9, Unit 1 was reduced to 88 percent RTP in response to fluctuating levels in the 6A and 6B feedwater heaters. Unit 1 returned to 100 percent RTP on November 10, 2013 and maintained approximately 100 percent RTP through the end of the report period.

Unit 2 maintained approximately 100 percent (RTP) during the report period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

a. Inspection Scope

Impending Adverse Weather Conditions The inspectors reviewed the licensees preparations to protect risk-significant systems from projected tropical storm force winds expected during October 4 - 6, 2013. The inspectors evaluated the licensees implementation of adverse weather preparation procedures and compensatory measures, including operator staffing, before the onset of the adverse weather conditions. The inspectors reviewed the licensees plans to address the ramifications of potentially lasting effects that may result from tropical storm force winds. The inspectors verified that operator actions specified in the licensees adverse weather procedure maintain readiness of essential systems. The inspectors also verified the licensee implemented periodic equipment walk-downs or other measures to ensure that the condition of plant equipment met operability requirements. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

1R04 Equipment Alignment

a. Inspection Scope

Partial Walk-Down The inspectors verified that critical portions of selected risk-significant systems were correctly aligned. The inspectors selected systems for assessment because they were a redundant or backup system/train, were important for mitigating risk for the current plant

conditions, had been recently realigned, or were a single-train system. The inspectors determined the correct system lineup by reviewing plant procedures and drawings. The inspectors verified that critical portions of the selected systems were correctly aligned by performing partial walkdowns. Documents reviewed are listed in the Attachment. The inspectors selected the three following system/trains to inspect:

• U1 A train residual heat removal (RHR) aligned for shutdown cooling

• U1 B train component cooling water (CCW) during maintenance on C CCW pump

• 1B emergency diesel generator

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Quarterly Inspection

The inspectors evaluated the adequacy of selected fire plans by comparing the fire plans to the defined hazards and defense-in-depth features specified in the fire protection program. In evaluating the fire plans, the inspectors assessed the following items:

1) control of transient combustibles and ignition sources; 2) fire detection systems; 3) water-based fire suppression systems; 4) gaseous fire suppression systems; 5) manual firefighting equipment and capability; 6) passive fire protection features; 7) compensatory measures and fire watches; and 8) issues related to fire protection contained in the licensees corrective action program. The inspectors toured the following four fire areas to assess material condition and operational status of fire protection equipment. Documents reviewed are listed in the Attachment.

• Unit 2, containment spray pump rooms, Fire Area 2-1

.2 Annual Inspection

On November 21, the inspectors evaluated the licensees fire brigade performance during a drill and assessed the brigades capability to meet fire protection requirements.

The inspectors observed the following aspects of fire brigade performance: 1) leadership ability of the brigade leader; 2) team effectiveness; and 3) compliance with site procedures. The inspectors also assessed the ability of control room operators to combat potential fires, including identifying the location of the fire, dispatching the fire brigade, and sounding alarms. The inspectors evaluated the licensees ability to declare the appropriate emergency action level and make required notifications in accordance with NUREG-0654 and 10 CFR 50.72. Documents reviewed are listed in the

.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

a. Inspection Scope

Internal Flooding The inspectors reviewed related flood analysis documents and walked down the area listed below that contain risk significant structures, systems, and components susceptible to flooding. The inspectors verified 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 their corrective action program.

Documents reviewed are listed in the attachment.

• Unit 1 auxiliary building, 100 foot elevation, auxiliary feedwater (AFW) pump rooms

b. Findings

No findings were identified.

1R08 Non-Destructive Examination Activities and Welding Activities

a. Inspection Scope

During the October 2013 station refueling outage the inspectors conducted an on-site containment walk down, and from December 16 -24, 2013, the inspectors conducted a review of documents associated with the implementation of the licensees in-service inspection (ISI) program for monitoring degradation of the reactor coolant system, emergency feedwater systems, risk-significant piping and components, and containment systems in Unit 1.

The inspectors activities included a review of non-destructive examinations (NDEs) to evaluate compliance with the applicable edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC),Section XI, and to verify that indications and defects were appropriately evaluated and dispositioned in accordance with the requirements of the ASME Code,Section XI, acceptance standards or NRC approved alternative requirement.

The inspectors reviewed records of the following NDE mandated by the ASME Code to evaluate compliance with the ASME Code Section XI and Section V requirements, and determine if any indications and defects were detected. Inspectors also reviewed evaluations of results that were dispositioned in accordance with the ASME Code or an NRC-approved alternative requirement.

The inspectors reviewed Ultrasonic (UT) examination records of the following Safety Injection System Welds:

• ALA1-4102-10 - Elbow to pipe weld

• ALA1-4102-11 - Pipe to elbow weld

• ALA1-4102-12 - Pipe to pipe weld The inspectors reviewed documentation for the repair/replacement of the following pressure boundary welds. The inspectors evaluated if the licensee applied the pre-service non-destructive examinations and acceptance criteria required by the construction code. In addition, the inspectors reviewed the welding procedure specifications, welder qualifications, welding material certifications, and supporting weld procedure qualification records to evaluate if the weld procedures were qualified in accordance with the requirements of construction code and the ASME Code Section IX.

• Work Order 1062195501 - Replacement of Seal Water Injection valve Q1E21V172A PWR Vessel Upper Head Penetration (VUHP) Inspection Activities Inspectors reviewed implementing procedures and the examination results of the Unit 1 vessel head bare metal visual (BMV) examination conducted in accordance with the requirements of ASME Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D).

Boric Acid Corrosion Control (BACC) Inspection Activities The inspectors reviewed the licensees BACC program activities to ensure implementation with commitments made in response to NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and applicable industry guidance documents. Specifically the inspectors performed a record review of procedures and the results of the licensees containment walkdown inspections performed during the current refueling outage.

The inspectors conducted an independent walkdown of containment to evaluate compliance with licensees BACC program requirements, and verified that degraded or non-conforming conditions, such as boric acid leaks, were properly identified and corrected in accordance with the licensees BACC and corrective action programs.

The inspectors reviewed the following evaluations and corrective actions related to evidence of boric acid leakage to evaluate if the corrective actions completed were consistent with the requirements of the ASME Code Section XI and 10 CFR Part 50, Appendix B, Criterion XVI.

• CR 437663 - Boron buildup on flow meter Q2E21F7904l

• CR 535915 - Boric acid leaking at valve Q2E21V165B

Steam Generator (SG) Tube Inspection Activities The licensee did not perform SG tube inspection activities during this outage. The inspectors reviewed the licensees Degradation Assessment to verify that the licensee met the requirements for skipping these inspections for this refueling maintenance cycle based on the licensees technical specifications, NRC commitments, ASME Section XI, and Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines.

Identification and Resolution of Problems The inspectors reviewed a sample of ISI-related problems that were identified by the licensee and entered into the corrective action program as condition reports (CRs). The inspectors reviewed the CRs to confirm the licensee had appropriately described the scope of the problem and had initiated corrective actions. The review also included the licensees consideration and assessment of operating experience events applicable to the plant. The inspectors performed this review to ensure compliance with 10CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requirements. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

.1 Resident Inspector Quarterly Review of Licensed Operator Requalification

The inspectors observed a simulator scenario conducted for training of an operating crew for continuing training on November 13, 2013. The inspectors assessed licensed operator performance, the ability of the licensee to administer the scenario and evaluate the operators, the quality of any post-scenario critique, any follow-up actions taken by the facility licensee, and the performance of the simulator. Documents reviewed are listed in the Attachment.

.2 Resident Inspector Quarterly Review (Licensed Operator Performance):

The inspectors observed licensed operator performance in the main control room during a Unit 1 Reactor Startup on October 25, 2013. Inspectors observed licensed operator performance to assess the following:

• Use of plant procedures

• Control board manipulations

• Communications between crew members

• Use and interpretation of instruments, indications, and alarms

• Use of human error prevention techniques

• Documentation of activities

• Management and supervision Documents reviewed are listed in the attachment.

.3 Annual Review of Licensee Requalification Examination Results

On May 7, 2013, the licensee completed the annual requalification operating examinations required to be administered to all licensed operators in accordance with 10 CFR 55.59(a)(2). The inspectors performed an in-office review of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with Inspection Procedure (IP) 71111.11, Licensed Operator Requalification Program. These results were compared to the thresholds established in Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Appendix I, Operator Requalification Human Performance Significance Determination Process.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors assessed the licensees treatment of the three issues listed below in order to verify the licensee appropriately addressed equipment problems within the scope of the Maintenance Rule (10 CFR 50.65). The inspectors reviewed procedures and records in order to evaluate the licensees identification, assessment, and characterization of the problems as well as their corrective actions for returning the equipment to a satisfactory condition. The inspectors also interviewed system engineers to assess the accuracy of equipment deficiencies and extent of condition. Documents reviewed are listed in the Attachment.

• Control Room Heating Ventilation and Air Conditioning System - (a)(2) performance criteria evaluation

• CR 713878, Local Leak Rate Testing (LLRT) failures on containment penetration 45 for excess letdown heat exchanger CCW inlet line

• CR 649708, Service Air Compressor tripped on low oil pressure

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the maintenance activities listed below to verity the licensee assessed and managed plant risk as required by 10 CFR 50.65(a)(4) and licensee procedures. The inspectors assessed the adequacy of the licensees risk assessments and implementation of risk management actions. The inspectors also verified that the licensee was identifying and resolving problems with assessing and managing maintenance-related risk using the corrective action program. Additionally, for maintenance resulting from unforeseen situations, the inspectors assessed the effectiveness of the licensees planning and control of emergent work activities.

Documents reviewed are listed in the attachment.

• Unit 1, October 1, yellow risk for shutdown safety assessment for RCS level drain down to below reactor vessel flange

• Unit 2, October 10, increased Green risk condition for high voltage switchyard maintenance activities

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors selected the four operability determinations or functionality evaluations listed below for review based on the risk-significance of the associated components and systems. The inspectors reviewed the technical adequacy of the determinations to ensure that technical specification operability was properly justified and the components or systems remained capable of performing their design functions. To verify whether components or systems were operable, the inspectors compared the operability and design criteria in the appropriate sections of the technical specification and updated final safety analysis report to the licensees evaluations. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. Additionally, the inspectors reviewed a sample of corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with operability evaluations.

Documents reviewed are listed in the attachment.

• CR 720475, Instrument air leak on unit 2 turbine driven auxiliary feedwater pump (TDAFWP) steam admission valve, HV3226

• CR 742443, 1B charging pump Mechanism Operated Cell (MOC) switch not cycling

• CR 743413, 1-2A EDG Essential Protection Relay 86A indication not lit

• CR 703677, Unit 1 auxiliary feedwater suction pipe support, HEG-548-AFW-H523 broken

b. Findings

No findings were identified.

1R18 Plant Modifications

a. Inspection Scope

The inspectors verified that the two plant modifications listed below did not affect the safety functions of important safety systems. The inspectors confirmed the modifications did not degrade the design bases, licensing bases and performance capability of risk significant structures, systems and components. The inspectors also verified modifications performed during plant configurations involving increased risk did not place the plant in an unsafe condition. Additionally, the inspectors evaluated whether system operability and availability, configuration control, post-installation test activities, and changes to documents, such as drawings, procedures, and operator training materials, complied with licensee standards and NRC requirements. In addition, the inspectors reviewed a sample of related corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with modifications. Documents reviewed are listed in the Attachment.

Permanent Plant Modifications

• SNC335478, Replacement of Unit 1 Turbine Driven Auxiliary Feedwater (AFW)

Pump Uninterruptible Power Supply (UPS)

• SNC528583, AFW Pipe Support HEG-548-AFW-H523 Redesign

b. Findings

No findings were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

The inspectors either observed post-maintenance testing or reviewed the test results for the four maintenance activities listed below to verify the work performed was completed correctly and the test activities were adequate to verify system operability and functional capability. The inspectors evaluated these activities for the following: acceptance criteria were clear and demonstrated operational readiness; effects of testing on the plant were adequately addressed; test instrumentation was appropriate; tests were performed in accordance with approved procedures; equipment was returned to its operational status following testing; and test documentation was properly evaluated. Additionally, the inspectors reviewed a sample of corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with post-maintenance testing.

Documents reviewed are listed in the Attachment.

• FNP-1-STP-40.0B, Safety Injection with Loss of Off-Site Power Test - B Train following replacement of 1B EDG output breaker

• FNP-1-STP-627.0, Local Leak Rate Testing of Containment Penetrations, following repairs on Q1P19V002

• FNP-1-STP-4.3, 1C Charging Pump Quarterly Inservice Test, following maintenance on minimum flow line check valve and pump handswitch replacement

• FNP-1-FSP-405.0, Preaction Sprinkler System Fire Surveillance, partial performance of Zone 1A-36

b. Findings

A licensee-identified violation was identified. The enforcement aspects are discussed in section 4OA7.

1R20 Refueling and Other Outage Activities

Unit 1 Scheduled Refueling Outage

a. Inspection Scope

The inspectors evaluated the outage activities listed below for the Unit 1 refueling outage from September 29, 2013 through October 27, 2013. The inspectors verified that the licensee: 1) considered risk in developing the outage schedule, 2) controlled plant configuration in accordance with administrative risk reduction methodologies, 3)developed work schedules to manage fatigue, 4) developed mitigation strategies for loss of key safety functions, and 5) adhered to operating license and technical specification requirements. Additionally, inspectors verified that safety-related and risk significant structures, systems, and components not accessible during power operations were maintained in an operable condition.

• Outage planning

• Cooldown, refueling, heatup, and startup

• Reactor coolant system instrumentation and electrical power configuration

• Reactivity and inventory control

• Decay heat removal and spent fuel pool cooling system operation

• Containment closure

• Local Leak Rate Testing (LLRT) of containment penetrations Additionally, the inspectors reviewed a sample of related corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with outage activities. Documents reviewed are listed in the attachment.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the surveillance test listed below and either observed the test or reviewed test results to verify testing adequately demonstrated equipment operability and met Technical Specification and licensee procedural requirements. The inspectors evaluated the test activities to assess for preconditioning of equipment, procedure adherence, and equipment alignment following completion of the surveillance.

Additionally, the inspectors reviewed a sample of related corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with surveillance testing. Documents reviewed are listed in the Attachment.

Routine Surveillance Tests

• FNP-1-STP-40.0B, Safety Injection with Loss of Off-Site Power Test

b. Findings

No findings were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The NSIR headquarters staff performed an in-office review of the latest revisions of various Emergency Plan Implementing Procedures (EPIPs) and the Emergency Plan located under ADAMS accession numbers ML130320653, ML13165A369, ML131680031, ML13191B308, and ML13214A050, as listed in the Attachment.

The licensee determined that in accordance with 10 CFR 50.54(q), the changes made in the revisions resulted in no reduction in the effectiveness of the Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, these revisions are subject to future inspection. The specific documents reviewed during this inspection are listed in the attachment. This inspection activity satisfied one inspection sample for the emergency action level and emergency plan changes on an annual basis.

b. Findings

No findings were identified.

4OA1 Performance Indicator Verification

a. Inspection Scope

The inspectors reviewed a sample of the performance indicator (PI) data, submitted by the licensee, for the PIs listed below. To verify the accuracy and completeness of the data reported for the station, the inspectors reviewed plant records compiled between December 2012 and December 2013. The inspections verified that the PI data complied with guidance contained in NEI 99-02, Regulatory Assessment Indicator Guideline, and licensee procedures. The inspectors also confirmed the PIs were calculated correctly.

In addition, the inspectors reviewed a sample of related corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with PI data. Documents reviewed are listed in the Attachment.

Cornerstone: Mitigating Systems

• High Pressure Injection System

• Cooling Water System

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

.1 Routine Review

The inspectors performed a daily screening of items entered into the licensees corrective action program in order to identify repetitive equipment failures or specific human performance issues for follow-up. The inspectors reviewed daily condition reports, attended screening meetings, or accessed the licensees computerized corrective action database.

.2 Semi-Annual Trend Review:

a. Inspection Scope

The inspectors reviewed the licensees corrective action program and associated documents to identify trends which could indicate the existence of a more significant safety issue. The inspectors focused their review on multiple issues associated with test control, but also considered the results of inspector daily condition report screenings, licensee trending efforts, and licensee human performance results. The review nominally considered the six month period of July 2013 through December 2013 although some examples extended beyond those dates when the scope of the trend warranted. The inspectors compared their results with the results contained in the licensees trend documents. Additionally, the inspectors reviewed the adequacy of

corrective actions associated with a sample of the issues identified in the licensees trend reports. The inspectors also reviewed corrective action documents which have been processed by the licensee to identify potential adverse trends in structures, systems, and/or components as evidenced by acceptance of long-standing non-conforming or degraded conditions. Documents reviewed are listed in the Attachment.

b. Findings

/Observations The inspectors identified a minor violation 10 CFR 50 Appendix B, Criterion XI, Test Control with three examples. 10 CFR 50 Appendix B, Criterion XI required in part that all testing required to demonstrate that structures, systems and components (SSCs) will perform satisfactorily in service is identified and performed in accordance with written test procedures. Contrary to the above, the licensee failed to follow the written test procedures as described below.

• On October 14, during the performance of FNP-1-STP-80.15 (EDG B Train LOSP Sequencer B1J Load Shedding Circuit and 2C DG SBO Start Test), the B1J sequencer load shed did not occur as expected. The sequencer did not function as expected because jumper 27XJ was erroneously left in the B1J sequencer from a previous STP-40.0B surveillance conducted ten days earlier. The performance deficiency is minor because the test was re-conducted and the error was observed and corrected before the B1J sequencer was placed back in service. The licensee wrote CR 717940 to address the issue.

• On October 16, during the performance of FNP-1-STP-40.0B (Safety Injection with LOSP Test, B Train) , the 1C Charging Pump breaker was incorrectly documented as RACKED TO TEST on Attachment 1 of the procedure. As a result, breaker DG06 did not close as expected after the actuation of the LOSP/ESF sequencer.

The performance deficiency is minor because the surveillance was re-conducted satisfactory on the charging pump breaker prior to returning it to service. The licensee wrote CR 718977 to address the issue.

• On October 18, while conducting LLRT of Pen 48 IAW FNP-1-STP-627.0 (Containment Local Leak Rate Test), the test was conducted with an improper system alignment as required by Attachment D of the procedure. The test was conducted with the vent valve downstream of check valve Q1P19V0002 erroneously left in the open position. This yielded an initial LLRT with satisfactory results. While restoring from the test, station personnel observed and corrected the lineup and conducted the test again, this time with unsatisfactory results. The performance deficiency is minor because the error was observed and the test results analyzed before the valve was placed back in service. The licensee wrote CR 720328 to address the issue.

The inspectors determined these performance deficiencies were indicative of an adverse trend in the licensees surveillance/test control program. The licensee documented the trend in their CAP as CR 762553.

.3 Annual Follow-up of Selected Samples

a. Inspection Scope

The inspectors selected the following two condition reports for detailed reviews:

• CR 6648187, Calculation error affects emergency action level (EAL) setpoints,

• CR 723304, Evaluate unfused battery ammeter circuits The inspectors evaluated the following attributes of the licensees actions:

• complete and accurate identification of the problem in a timely manner

• evaluation and disposition of operability/reportability issues

• consideration of extent of condition, generic implications, common cause, and previous occurrences

• classification and prioritization of the problem

• identification of root and contributing causes of the problem

• identification of any additional condition reports

• completion of corrective actions in a timely manner Documents reviewed are listed in the attachment.

.1 Calculation Error Results in Significantly non-Conservative EAL Threshold Values

Introduction:

A White finding and associated violation of 10 CFR 50.54(q)(2) was identified by the licensee for the failure to follow and maintain the effectiveness of emergency plans which meet the requirements of 10 CFR 50.47(b)(4). Specifically, the licensee's emergency classification scheme action levels for Category R - Abnormal Radiological General Emergency action level (EAL) RG1 and Site Area Emergency EAL RS1 contained declaration threshold values which were significantly higher than appropriate due to a calculation error.

Description:

In December 2004, a Southern Co. corporate engineering calculation, SM-96-1076-002, was developed to estimate dose rates as a function of radiological releases correlated to radiation monitor values. The calculation provided radiation monitor threshold values for General Emergency (i.e. exceeding 1000 mRem TEDE/5000 mRem thyroid CDE beyond the site boundary) and Site Area Emergency (i.e. exceeding 100 mR TEDE/500 mRem thyroid CDE beyond the site boundary). The calculation was a manual calculation using a spreadsheet program; however, a unit conversion error (Sieverts/second to mRem/hour) was made and not detected during the review process. The error resulted in threshold values sixty times greater than the desired value. In 2005, Farley Nuclear Plant submitted a license amendment request to the NRC to change their EAL scheme to one based on NEI-99-01 Rev. 4 guidelines.

The request included EAL threshold values for RG1 and RS1 which were based on the errant calculation. The NRC approved the amendment and the licensee implemented

the EAL scheme by issuing revision 48 of the Farley Nuclear Plant emergency plan on September 24, 2010. The non-conservative threshold values were contained in the emergency plan.

During an extent of condition review prompted by Southern Co. fleet operating experience, calculation SM-96-1076-002 was reviewed and the calculation error was discovered. On May 31, 2013, the issue was placed in the licensees corrective action program as CR 648187. That same day, immediate corrective actions included providing corrected threshold values to appropriate management and decision-makers (shift managers/emergency directors). Additional actions taken included; performing an Enhanced Apparent Cause Determination per the licensees procedures, processing formal changes to the station emergency plan and associated implementing procedures, and performing extent of condition/cause reviews throughout the Southern Co. fleet.

NRC regional inspectors were advised of the issue and intended plan-of-action.

Following extensive review, the revised emergency plan and associated implementing procedure were provided to the NRC in July 2013.

These discrepant threshold values degraded the licensees ability to make timely and accurate General Emergency and Site Area Emergency classifications based on the Abnormal Radiological initiating condition, in that decision-makers would have to rely on other means to classify the event (dose assessments or field monitoring data) and that could delay such a declaration.

Analysis:

The inspectors concluded that the failure to maintain the effectiveness of an emergency plan to meet the requirements of 10 CFR 50.47(b)(4) and Part 50 Appendix E to have a standardized EAL scheme in use based on facility system and effluent parameters, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the emergency preparedness cornerstone attribute of procedure quality. It impacted the cornerstone objective because it was associated with inappropriate EAL and emergency plan changes and their adequacy to protect the health and safety of the public in the event of a radiological emergency. Specifically, the licensees ability to declare a Site Area and General Emergency based on effluent radiation monitor values was degraded in that event classification using these radiation monitors would be delayed. The finding was assessed for significance in accordance with NRC Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, which states, FAILURE TO COMPLY means that a program is noncompliant with a REGULATORY REQUIREMENT. The inspector determined the licensee was noncompliant with 10 CFR 50.54(q), 50.47(b)(4), and Appendix E,Section IV.B in that, due to a calculation error, the abnormal radiological initiating conditions RG1(General Emergency) and RS1 (Site Area Emergency) emergency action levels contained classification threshold values sixty times greater than the appropriate value. This would require use of other means (dose assessment or actual field readings) to determine whether a Site Area Emergency or General Emergency threshold had been exceeded which could delay the declaration.

The inspector determined that the situation constituted a degraded rather than failed risk-significant planning standard (RSPS). The issue of concern was similar to the example in Table 5.4.1 (Degraded RSPS) and was determined to be of low to moderate safety significance (White). The licensee took immediate corrective actions providing

corrected threshold values to appropriate management and decision-makers (shift managers/emergency directors). These and additional corrective actions were placed in the licensees corrective action process as CR 648187. A cross-cutting aspect was not assigned based on the elapsed time since the performance deficiency occurred and because the inspectors determined it was not reflective of current licensee performance.

Enforcement:

10 CFR 50.54(q)(2), requires that a holder of a nuclear power reactor operating license under this part, shall follow and maintain the effectiveness of emergency plans which meet the standards in 10 CFR 50.47(b), and the requirements in appendix E of this part.

10 CFR 50.47(b)(4), requires a standard emergency classification and action level scheme, the bases of which include facility and system effluent parameters is in use by the nuclear facility licensee, and State and local response calls for reliance on information by facility licensees for determinations of minimum initial offsite response measures.

10 CFR Part 50, Appendix E, Section IV.B., Assessment Actions, requires that means to be used for determining the magnitude of, and for continuously assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies.

The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring.

Contrary to the above, from September 2010 to May 2013, the licensee failed to maintain the effectiveness of its emergency plan. The licensee failed to maintain a standard emergency classification scheme which included facility effluent parameters.

Specifically, the emergency classifications RG1 (General Emergency) and RS1 (Site Area Emergency) contained effluent radiation monitor threshold values significantly greater than appropriate. These monitors were being relied upon to determine the magnitude and for continuously assessing the impact of the release of radioactive materials, as well as providing criteria for determining the need for notification and participation of local and State agencies. Following review by a Significance Enforcement Review Panel and NRC management, the violation was determined to meet IMC 0305, Section 06.06, criteria for enforcement discretion as an old design issue. Specifically, the issue was licensee-identified through an extent-of-condition review of internal operating experience, the issue was immediately corrected by the licensee, the issue was not likely to be previously identified during normal operations, routine testing, or maintenance, and the issue is not reflective of current licensee performance. As such, this finding will not be used as an input in the assessment process or NRC Action Matrix. This finding has been identified as a cited violation 05000348,364/2013005-01, Calculation Error Results in Significantly non-Conservative EAL Threshold Values. This is a violation of 10 CFR 50.54(q)(2) and a Notice of Violation is enclosed. (Enclosure 2)

.2 Evaluation of Unfused Battery Ammeter Circuits

A licensee-identified violation was identified for the issue associated with CR 723304.

The enforcement aspects are discussed in section 4OA7.

4OA3 Follow-up of Events

.1 (Closed) Licensee Event Report (LER) 05000348/2013-001-00, Automatic Reactor Trip

and B-Train Loss of Off-site Power Caused by the Failure of a Startup Transformer Lightning Arrester

a. Inspection Scope

The inspectors reviewed the LER described above, the associated root cause report (CAR 207147), the apparent cause determination report (CAR 207586) and discussed the issue with licensee staff. The licensee determined the direct cause of the Unit 1 reactor trip on June 11, 2013, was due to the loss of the 1B start-up transformer (SUT)and subsequent loss of the B and C reactor coolant pumps. The 1B SUT tripped due to failure of the phase 2 lightning arrester caused by moisture intrusion which resulted from a manufacturing deficiency of the arrester sealing plate. All three lightning arrestors were replaced and 1B SUT satisfactorily tested before being returned to service.

b. Findings

No findings were identified.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security.

These observations took place during both normal and off-normal plant working hours.

These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors normal plant status review and inspection activities.

b. Findings

No findings were identified.

.2 Operation of an Independent Spent Fuel Storage Installation (ISFSI) (IP 60855.1)

a. Inspection Scope

The inspectors performed a walkdown of the ISFSI on site and observed licensee personnel inspecting the ISFSI in accordance with plant procedures. The inspectors reviewed surveillance records to verify that daily surveillance requirements were performed as required by technical specifications. Documents reviewed are listed in the

.

b. Findings

No findings were identified.

.3 Follow-up Inspection for Confirmatory Order, EA-12-145, May 6, 2013, Failure to Ensure

that Radiation Worker Training (RWT) Exams for Security Officers Were Not Compromised.

a. Inspection Scope

The Nuclear Regulatory Commission (NRC) performed this inspection in accordance with Inspection Procedure (IP) 92702, Follow-up on Traditional Enforcement Actions Including Violations, Deviations, Confirmatory Action Letters, Confirmatory Orders and Alternate Dispute Resolution Confirmatory Orders, to assess the licensees corrective actions for Confirmatory Order EA-12-245 (ML13127A136.) The objective of the inspection was to review the adequacy of the implementation of commitments that were part of the Confirmatory Order. The Confirmatory Order was issued as a result of a successful Alternate Dispute Resolution (ADR). The commitments were made by Southern Nuclear Company (SNC) as part of the settlement agreement reached during the ADR session involving the failure to ensure that radiation worker training exams for security officers were not compromised. SNC - Farley Nuclear Plant (FNP) agreed to corrective actions and enhancements related to security test taking environment and proctoring of examinations fleet-wide and also the use of SNCs internal programs for resolution of issues/deficiencies.

The inspection was completed at SNC corporate office in Birmingham, Alabama. During the inspection, the inspectors evaluated the licensees response to the commitments in terms of adequacy of the licensees response to each commitment and completion within the times specified. The evaluation was conducted through interviews with licensee staff at the corporate office and an evaluation of licensee documents and procedures related to compliance with the Order.

b. Findings

/Observations The inspectors evaluated the completion and adequacy of the licensees response to Confirmatory Order EA-12-145. The inspectors reviewed the current status of each commitment in Paragraph V. A summary of the licensees response to each commitment is included below:

• SNC assessed the results of the completed evaluation of the testing environment for the Security Departments at all three sites. SNC determined no further actions were necessary at FNP. Corrective actions were submitted for Hatch Nuclear Plant (HNP)and Vogtle Nuclear Plant (VNP) to add Security Management observations for examinations by security officers. This commitment is considered closed.

(Paragraph V.a)

• SNC evaluated the testing environment and compliance with NMP-TR-208, Examination and Examination Security, at the corporate offices and the three operating sites. SNC assessed the results of the evaluations for the testing environment fleet-wide and compliance with NMP-TR-208. As a result a corrective action, CR 663661, was created and the lesson plan for Department Training Coordinator, course number S-GE-96701 Version 2.1, was updated to include additional guidance from NMP-TR-208 related to exam security. This commitment is considered closed. (Paragraph V.b)

• By June 15, 2013, SNC had reviewed listed procedures and sub-tier documents referenced in NMP-TR-208 related to proctoring. They determined the procedures were in alignment as it relates to proctoring and no additional changes were needed.

This commitment is considered closed. (Section V.c)

• SNC issued a fleet-wide communication, on June 7, 2013, regarding the revisions and clarifications that had been made to NMP-TR-208 and other procedures referenced in NMP-TR-208. The communication was included in a fleet-wide email News You Can Use, with an additional section listing company expectations and requirements for employees regarding procedure NMP-TR-208. This commitment is considered closed. (Paragraph V.d)

• A fleet wide communication in the form of a video with a message from Danny Bost, SNC Chief Nuclear Officer, was sent mid-June 2013 to all available employees. A list of employees that were unavailable was entered into the corrective action program to ensure they received the communication. SNC is tracking a minimal number of employees who upon first availability will be required to view the video.

The message clearly articulated that willful misconduct is incompatible with safe nuclear construction and operation. The communication provided recent examples, including those documented in EA-12-240, Notice of Violation regarding missed fire watches, and EA-12-230, Licensee Identified Violation regarding falsified construction check forms, and the impacts of loss of integrity and trustworthiness.

This commitment is considered closed. (Paragraph V.e)

• By July 30, 2013, following the senior management communication to address integrity and trustworthiness during a fleet-wide stand-down was conducted with all available employees. A minimal list of unavailable employees was entered into the corrective actions program to ensure they receive the communication upon their

return to the site. Vogtle 3 and 4 SNC executive level leadership performed a stand down with both contractors and employees. This commitment is considered closed.

(Paragraph V.f)

• A new procedure NL-008, Response to NRC-Referred Allegations and Investigations, Version 1.0 approved July 29, 2013, was created to provide guidance involving investigations based on allegations. The procedure includes an initial evaluation for potential nuclear safety implications and an evaluation to identify any immediate mitigating measures. This commitment is considered closed.

(Paragraph V.g)

• An effectiveness review of all corrective actions taken under the Confirmatory Order is due by August 31, 2014. This commitment remains open. (Paragraph V.h)

• SNC planned to reinforce the messages from paragraphs V.e and V.f annually through 2015. They have assigned tasks with completion dates through their corrective action program to ensure this commitment is met. This commitment is considered closed. (Paragraph V.i)

• Upon completion of the terms of paragraph V.a through V.h as directed by the Confirmatory Order, SNC will provide the NRC with a letter discussing the basis for concluding that the order has been satisfied. This commitment remains open.

(Paragraph V.j)

The closure of Confirmatory Order EA-12-145 is pending the results of the effectiveness review and a letter to the NRC from SNC discussing its basis for concluding that the order has been satisfied.

No findings were identified.

4OA6 Meetings, Including Exit

The NRC presented the inspection results to Cheryl Gayheart, Site Vice-president and members of the licensees staff on January 30, 2014. The staff acknowledged the results. The NRC confirmed that any proprietary information that was provided during the inspection period was properly controlled or returned.

4OA7 Licensee-Identified Violations

The following violations of very low safety 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 a Non-Cited Violations.

• 10 CFR 50.72(b)(3)(ii)(B), Immediate notification requirements for operating nuclear power reactors -

(3) Eight-hour reports, required the licensee to notify the NRC as soon as practical and in all cases within eight hours of the occurrence of the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

Contrary to the above, the licensee failed to make the required notification within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of an issue associated with certain unfused direct current (DC) ammeters in the main control room. CR 476438 was written on June 27, 2012, which describes the potential fire vulnerability to alternate shutdown capability. Technical evaluation (TE) 451064, completed on July 12, 2012, confirmed the vulnerability in that the ammeter circuits do not contain fuses which would provide overcurrent protection.

TE 449432 was completed on August 21, 2012 and incorrectly concluded that this issue was not reportable. CR 723304 was written on October 24, 2013, to re-evaluate the reportability requirements of this issue. The licensee notified the NRC of this issue on December 16, 2013, via event notification (EN) 49638. Since this finding impacted the ability of the NRC to perform its regulatory oversight function, it was evaluated using the traditional enforcement process. The inspectors concluded that failure to make the required notification within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was a Severity Level IV violation in accordance with Section 6.9(d) of the NRCs Enforcement Policy.

• Technical Specifications 5.4.1.c, Fire Protection Program Implementation, required in part that written procedures be established, implemented, and maintained covering activities of the fire protection program. Contrary to the above requirements, the licensee failed to establish an adequate written procedure (work order SNC 505601) covering activities of the fire protection program. Specifically, on December 10, 2013, per WO SNC 505601, the licensee replaced a degraded fire indication unit (FIU) that contained a 15-ampere (A) fuse on the Unit 1 pyro panel with an FIU that contained a 5A fuse. On December 16, 2013, during performance of fire surveillance FNP-1-FSP-307.0, Smoke Detector - Biennial Operability and Adjustment, the 5A fuse opened due to an over-current condition, disabling the Unit 1 pyro panel and resulting in the loss of automatic fire detection capability in the Unit 1 auxiliary building. The licensee initiated CR 745702 in response to this issue and established the necessary fire watches in the Unit 1 auxiliary building until the pyro panel was returned to service on December 17, 2013. The inspectors screened the finding in accordance with IMC 0609 Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, and concluded the finding did not impact the ability of Unit 1 to achieve safe shutdown and was considered to be of very low safety significance (Green).

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

B. Arens, Licensing Supervisor

H. Cooper, Engineering Programs Supervisor

D. Drawbaugh, EP Manager

D. Enfinger, Corrective Action Program Supervisor

C. Gayheart, Plant Manager, Site Vice President

D. Guthrie, Corporate Security Manager

D. Hobson, Shift Operations Manager

L. Hogg, Fleet Security Project Manager

J. Horn, Regulatory Affairs Manager

J. Hutto, Engineering Director, Plant Manager

P. Ivey, Regulatory Affairs VP

M. Long, Fleet Security Manager

T. Lynch, Site Vice President

R. Martin, Engineering Programs Manager

S. McGavin, Fleet Security Director

D. McKinney, Licensing Manager

C. Pierce, Regulatory Affairs Director

B. Reed, Nuclear Operations Training Supervisor

D. Reed, Operations Support Manager

L. Riley, Performance Improvement

I. Sarygin, Sr. Engineer

D. Simmons, EP Specialist

B. Taylor, Nuclear Oversight Supervisor

C. Thornell, Operations Director

C. Westberry, Engineering Programs Supervisor

L. Williford, Licensing Engineer

NRC personnel

Frank Ehrhardt, Chief, Branch 2, Division of Reactor Projects

LIST OF ITEMS

OPENED AND CLOSED

Opened and Closed

None

Opened

05000348-364/2013005-01 VIO Calculation Error Results in Significantly non-

Conservative EAL Threshold Values (Section 4OA2.3)

Closed

05000348/2013-001-00 LER Automatic Reactor Trip and B-Train Loss of Offsite Power Caused by the Failure of a Startup Transformer Lightning Arrester (Section 4OA3.1)

Discussed

EA-12-145 CO Failure to Ensure that Radiation Worker Training (RWT) Exams for Security Officers Were Not Compromised (Section 4OA5)