IR 05000424/2014003

IR 05000424-14-003 and 05000425-14-003, and Notice of Violation, April 1, 2014, Through June 30, 2014, Vogtle Electric Generating Plant, NRC Integrated Report

ML14218A669
Person / Time
Site:	Vogtle
Issue date:	08/06/2014
From:	Munday J T Division Reactor Projects II
To:	Madison D Southern Nuclear Operating Co
References
EA-14-112 IR-14-003
Download: ML14218A669 (50)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION II

245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 August 6, 2014

EA-14-112 Mr. Dennis Madison

Vice President - Vogtle

Southern Nuclear Operating Company, Inc.

Vogtle Electric Generating Plant

7821 River Road

Waynesboro, GA 30830

SUBJECT: VOGTLE ELECTRIC GENERATING PLANT - NRC INTEGRATED INSPECTION REPORT 05000424/2014003 AND 05000425/2014003, AND NOTICE OF

VIOLATION

Dear Mr. Madison:

On June 30, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at

your Vogtle Electric Generating Plant, Units 1 and 2. On July 25, 2014, the NRC inspectors

discussed the results of this inspection with Mr. Tom Tynan and other members of the Vogtle

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 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 a potential 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 concern. The licensee took additional corrective

actions, including performing a causal determination, processing formal changes to the station's

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 a telephone conversation on July 3, 2014, Mr. Brian Bonser, Chief, Plant Support Branch, Division of Reactor Safety, Region II, informed Mr. Tynan of the details of the preliminary

finding, the apparent violation, and advised Vogtle representatives that the finding satisfied 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 Mr. Tynan that

based on the above, the NRC had sufficient information, including Vogtle's corrective actions, to

make a final significance determination and enforcement decision without the need for a

regulatory conference or a written response from you. Mr. Tynan indicated they 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 NRC's 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,"

to review the licensee's root cause and corrective action plans even if the White finding meets

the criteria for treatment as an old design issue. 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.

NRC inspectors also documented three findings of very low safety significance (Green)

identified during this inspection period. These findings involved violations of NRC requirements.

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 U.S. 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

Vogtle Electric Generating 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

Vogtle Electric Generating Plant.

In accordance with Title 10 of the Code of Federal Regulations 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC's "Agency Rules of Practice and

Procedure," a copy of this letter, its enclosu res, and your response (if any) will be available electronically for public inspection in the NRC P ublic Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Document Access and

Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/ Joel T. Munday, Director

Division of Reactor Projects

Docket Nos.: 05000424, 05000425 License Nos.: NPF-68 and NPF-81

Enclosures:

1. Inspection Report 05000424/2014003 and 05000425/2014003

w/Attachment: Supplemental Information 2. Notice of Violation

cc Distribution via ListServ

_____ ________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS RII:DRS RII:DRS SIGNATURE Via email Via email Via email Via email Via email Via email Via email NAME MCain TChandler WPursley A Nielsen WLoo CDykes MSpeck DATE 7/11/2014 7/24/2014 7/14/2014 7/14/2014 7/24/2014 7/24/2014 7/15/2014 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICE RII:DRS RII:DRS RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP SIGNATURE Via email Via email JGW /RA/ AXA /RA/ FJE /RA/ MSL /RA/ JTM /RA/ NAME SSanchez GOttenberg JWorosilo AAlen FEhrhardt MLesser JMunday DATE 7/14/2014 7/14/2014 7/22/2014 7/22/2014 8/4/2014 8/4/2014 8/6/2014 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICE RII:EICS SIGNATURE SAP /RA/ NAME SPrice DATE 8/1/2014 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO Letter to Dennis Madison from Joel T. Munday dated August 6, 2014.

SUBJECT: VOGTLE ELECTRIC GENERATING PLANT - NRC INTEGRATED INSPECTION REPORT 05000424/2014003 AND 05000425/2014003, AND NOTICE OF

VIOLATION

DISTRIBUTION

D. Gamberoni, RII

L. Douglas, RII

OE Mail RIDSNRRDIRS

PUBLIC RidsNrrPMVogtle Resource

Enclosure 1 U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos.: 50-424, 50-425

License Nos.: NPF-68, NPF-81

Report Nos.: 05000424/2014003 and 05000425/2014003

Licensee: Southern Nuclear Operating Company, Inc. (SNC)

Facility: Vogtle Electric Generating Plant, Units 1 and 2

Location: Waynesboro, GA 30830

Dates: April 1, 2014, through June 30, 2014

Inspectors: M. Cain, Senior Resident Inspector T. Chandler, Resident Inspector A. Alen, Project Engineer W. Pursley, Health Physics Inspector (2RS1, 2RS2, 2RS4, 4OA1) A. Nielsen, Senior Health Physicist (2RS1, 2RS3, 4OA1)

W. Loo, Senior Health Physicist (2RS1, 2RS3)

C. Dykes, Health Physicist (2RS5) M. Speck, Senior Emergency Preparedness Inspector (4OA2.3) S. Sanchez, Senior Emergency Preparedness Inspector

(4OA2.3) G. Ottenberg, Senior Reactor Inspector (4OA5)

Approved by: Frank Ehrhardt, Chief Reactor Projects Branch 2

Division of Reactor Projects Enclosure 1

SUMMARY OF FINDINGS

IR 05000424/2014003, 05000425/2014003; 04/01/2014 - 06/30/2014; Vogtle Electric

Generating Plant, Units 1 and 2; Maintenance Effectiveness, Radiological Hazard

Assessment and Exposure Controls, Identification and Resolution of Problems, Event

Follow-up The report covered a 3-month period of inspection by resident inspectors and regional inspectors. There was one NRC-identified and three self-revealing violations identified and 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. The cross-cutting aspects are determined using IMC 0310,

"Aspects within the Cross-Cutting Areas" dated December 19, 2013. All violations of

NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy dated January 28, 2013. The NRC's program for overseeing the safe operations of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight

Process," Revision 5.

Cornerstone: Initiating Events

Green A self-revealing non-cited violation (NCV) of 10 CFR 50 Appendix B Criterion V, "Instructions, Procedures, and Drawings," was identified for failure to provide adequate work instructions in the maintenance procedure used for main steam isolation valve (MSIV) maintenance. Specifically, maintenance procedure 26854-C, "Main Steam

Isolation Valve Actuator Maintenance," used to perform maintenance on Rockwell

MSIV(s), did not provide adequate instructions for installing the lower manifold/cylinder

O-ring during reassembly. This resulted in a 'pinched' O-ring on 1HV3006B, a subsequent failure of the O-ring causing the MSIV to fail closed, and a manual reactor trip. The licensee conducted a root cause investigation and entered the event into their corrective action program (condition report (CR) 800018). The licensee replaced the O-

ring, performed an extent of condition evaluation for all other MSIVs, and revised the maintenance procedure to include specific instructions for the installation of the lower manifold/cylinder O-ring.

The finding was more than minor because it was associated with the procedure quality attribute of the reactor safety - initiating events cornerstone and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.

Specifically, the failure to provide an adequate procedure with adequate instructions for ensuring proper O-ring installation resulted in the failure of the Unit 1 loop 1 outboard

MSIV hydraulic actuator causing the loop 1 MSIV to fail closed and a subsequent manual reactor trip due to lowering steam generator water level. Because the inspectors answered "No" to all of the IMC 0609 Appendix A (dated June 19, 2012) Exhibit 1,

Section B, "Initiating Events Screening Questions," the inspectors concluded that the finding was of very low safety significance (Green). The inspectors determined the finding had a cross-cutting aspect of "resources" in the human performance area, because the maintenance procedure used to install manifold/cylinder O-ring did not provide adequate instructions for the proper installation of the O-ring. [H.1] (Section 1R12)

Green A self-revealing NCV of 10 CFR 50 Appendix B Criterion V, "Instructions, Procedures, and Drawings," was identified for failure to provide adequate work instructions as well as failure to follow the maintenance procedure used to install flexible and rigid conduit. Specifically, the work instructions did not provide adequate directions and/or precautions to properly slope conduit during installation to prevent water intrusion into a valve positioner. The work instructions referenced maintenance procedure 25008-

C, "Flexible and Rigid Conduit Installation." The maintenance procedure referenced Vogtle design specification X3AR01 Section E-8, "Raceway Systems," which provided sloping and tightness criteria for conduit installations. The licensee conducted a root cause investigation and entered the event into their corrective action program (CR 797929). The licensee repaired the improperly sloped conduit, replaced the positioner, and revised procedure 25008-C to specify standards for proper sloping of conduits.

The finding was more than minor because it was associated with the procedure quality and human performance attributes of the reactor safety - initiating events cornerstone and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to provide adequate work instructions as well as failure to follow procedure 25008-C, "Flexible and Rigid Conduit Installation," resulted in the Unit 2 loop 3 main feedwater regulating valve (MFRV) positioner failing closed, causing a subsequent automatic reactor trip due to low-low steam generator (SG) water level. Because the inspectors answered "No" to all of the IMC 0609 Appendix A (dated

June 19, 2012) Exhibit 1, Section B, "Initiating Events Screening Questions," the inspectors concluded that the finding was of very low safety significance (Green). The inspectors determined that the finding had a cross-cutting aspect of "procedure adherence" in the human performance area because the maintenance electricians did not follow Vogtle design specification procedures or drawings resulting in the improper sloping of the MFRV flexible conduit [H.8] (Section 40A3)

Cornerstone: Occupational Radiation Safety

Green A self-revealing NCV of Technical Specification (TS) 5.7.1, "High Radiation Area", was identified for an entry into a high radiation area (HRA) without meeting the entry requirements as specified therein. Specifically, on March 17, 2014, an operator was authorized to enter an HRA on Unit 1 under conditions where dose rates were known to be changing. This allowed the operator entry into an HRA without knowledge of actual radiological conditions. He was not provided with a radiation monitoring device that continuously indicated dose rates in the area, nor was he accompanied by an individual qualified in radiation protection procedures with a radiation monitoring device providing positive control over his activities. Upon discovery of the condition, the licensee secured access to the area, performed follow-up surveys and convened a human performance review board to examine causal factors and identify corrective actions. The licensee entered this issue into the corrective action program as CR 787908. This finding was more than minor because it was associated with the occupational radiation safety cornerstone attribute of human performance and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, workers permitted entry into HRAs with inadequate knowledge of current radiological conditions could receive unintended occupational exposures. The finding was evaluated using IMC 0609, Appendix C, "Occupational Radiation Safety

Significance Determination Process (SDP)", dated August 19, 2008. The finding was not related to As Low As Reasonably Achievable (ALARA) planning, nor did it involve an overexposure or substantial potential for overexposure and the ability to assess dose was not compromised. Therefore, the finding was determined to be of very low safety significance (Green). This finding had a cross-cutting aspect of "avoid complacency" in the human performance area because health physics (HP) personnel failed to verify plant conditions through available means when an evolution was in progress that was known to increase area dose rates prior to authorizing entry into an HRA. [H.12] (Section 2RS1)

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 CR 648248.

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 licensee's ability to declare a Site Area Emergency 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 that the issue of concern constituted a degraded rather than lost 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)

REPORT DETAILS

Summary of Plant Status

Unit 1 started the reporting period shut down for a planned refueling outage. Operators

restarted the unit on April 11, 2014, and attained 100 percent rated thermal power (RTP) on

April 12, 2014. Operators manually tripped the unit on April 12, 2014, due to a failure of the

loop 1 main steam isolation valve (MSIV) failing closed at 100 percent RTP. Operators

restarted the unit on April 13, 2014 and attained 100 percent RTP on April 27, 2014. The unit

operated at essentially RTP for the rest of the inspection period.

Unit 2 started the report period at full RTP. The unit automatically tripped from 100 percent RTP

on April 8, 2014, due to low level in the loop 3 steam generator caused by the main feedwater

regulator valve (MFRV) failing closed. Operators restarted the unit on April 10, 2014, and

attained 100 percent power on April 11, 2014.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

.1 Summer Readiness of Offsite and Alternate AC Power System

Because the licensee implemented modifications to the high and low voltage

switchyards, the inspectors reviewed the licensee's procedures for operation and

continued availability of offsite and onsite alternate AC power systems. The inspectors

also reviewed the communications protocols between the transmission system operator

and the licensee to verify that the appropriate information is exchanged when issues

arise that could affect the offsite power system.

The inspectors reviewed the material condition of offsite and onsite alternate AC power

systems (including switchyard and transform ers) by performing a walkdown of the switchyard. The inspectors reviewed outstanding work orders and assessed corrective

actions for any degraded conditions that impacted plant risk or required compensatory

actions. Documents reviewed are listed in the Attachment.

.2 Seasonal Extreme Weather Conditions

The inspectors conducted a detailed review of the station's adverse weather procedures

written for extreme high temperatures. The inspectors verified that weather related

equipment deficiencies identified during the previous year had been placed into the work

control process and/or corrected before the onset of seasonal extremes. The inspectors

evaluated the licensee's implementation of adverse weather preparation procedures and compensatory measures before the onset of seasonal extreme weather conditions.

Documents reviewed are listed in the Attachment.

The inspectors evaluated the following risk-significant systems:

• Unit 2 nuclear service cooling water (NSCW) system (both trains)

• Unit 1 emergency diesel generator (EDG) system (both trains)

b. Findings

No findings were identified.

1R04 Equipment Alignment

a. Inspection Scope

Partial Walkdown

The inspectors verified that critical portions of the selected systems were correctly

aligned by performing partial walkdowns. The inspectors selected systems for

assessment because they were a redundant or ba ckup system or train, were important for mitigating risk for the current plant conditions, had been recently realigned, or were a

single-train system. The inspectors determi ned the correct system lineup by reviewing plant procedures and drawings. Documents reviewed are listed in the Attachment.

The inspectors selected the following four systems or trains to inspect:

• Unit 2 train "B" EDG while the train "A" EDG was out of service due to a planned maintenance outage

• Unit 2 train "B" motor-driven auxiliary f eedwater system and the train "C" turbine-driven auxiliary feedwater system duri ng the train "A" EDG planned maintenance outage

• Unit 2 train "A" EDG during the train "B" EDG planned maintenance outage

• Unit 2 train "A" motor-driven auxiliary f eedwater system and the train "C" turbine-driven auxiliary feedwater system duri ng the train "B" EDG planned maintenance outage

b. Findings

No findings were identified.

1R05 Fire Protection

a. Inspection Scope

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:

• control of transient combustibles and ignition sources

• fire detection systems

• water-based fire suppression systems

• gaseous fire suppression systems

• manual firefighting equipment and capability

• passive fire protection features

• compensatory measures and fire watches

• issues related to fire protection contained in the licensee's corrective action program

The inspectors toured the following five fire areas to assess material condition and

operational status of fire protection equipment. Documents reviewed are listed in the

.

• Unit 2 component cooling water (CCW) heat exchanger rooms, fire zones 54, 55, 148, 23, 172, and 147

• Unit 1 centrifugal charging pump (CCP) rooms and the level "C" pipe penetration area in the Unit 1 auxiliary building, fire zones 14B, 19, 20, and 21

• Unit 2 control building level "A" west and east penetration areas, fire zones 87, 88, 89, 90 93, 102, 158 and 159.

• Unit 1 "B" train EDG building, fire zones 162 and 164

• Unit 2 auxiliary feedwater pump house, fire zones 155, 156, 157A and 157B

b. Findings

No findings were identified.

1R06 Flood Protection Measures

.1 Internal Flooding

The inspectors reviewed related flood analysis documents and walked down the area

listed below containing risk-significant stru ctures, 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 correct ive action program. Documents reviewed are listed in the Attachment.

• Unit 1 residual heat removal (RHR) and containment spray (CS) pump rooms (both trains) in auxiliary building 1 1R11 Licensed Operator Requalification Pr ogram and Licensed Operator Performance (71111.11)

.1 Resident Inspector Quarterly Review of Licensed Operator Requalification

The inspectors observed an evaluated simulator scenario administered to an operating

crew conducted in accordance with the licensee's accredited requalification training

program.

The inspectors assessed the following:

• licensed operator performance

• the ability of the licensee to administer the scenario and evaluate the operators

• the quality of the post-scenario critique

• simulator performance Documents reviewed are listed in the Attachment.

.2 Resident Inspector Quarterly Review of Licensed Operator Performance

The inspectors observed licensed operator performance in the main control room on

April 9, 2014, while operators were starting up the Unit 2 reactor.

The inspectors assessed 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.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors assessed the licensee's treatment of the two issues listed below in order

to verify the licensee appropriately addressed equipment problems within the scope of the maintenance rule (10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants".) The inspectors reviewed procedures and

records in order to evaluate the licensee's identification, assessment, and

characterization of the problems as well as their corrective actions for returning the

equipment to a satisfactory condition. The in spectors also interviewed system engineers and the maintenance rule coordinator to assess the accuracy of performance

deficiencies and extent of condition. Documents reviewed are listed in the Attachment.

• Unit 2, system 1305, 2HV5230 hydraulic leak

• Unit 1, system 1301, 1HV3006B maintenanc e preventable functional failure (MPFF)

b. Findings

Introduction

A Green, self-revealing NCV of 10 CFR 50 Appendix B Criterion V, "Instructions, Procedures, and Drawings," was identified for failure to provide adequate

work instructions in the maintenance procedure used to reassemble Rockwell MSIVs.

Specifically, maintenance procedure 26854-C, "Main Steam Isolation Valve Actuator

Maintenance," which is used to perform maintenance on Rockwell MSIVs, did not

provide adequate instructions for installing of the lower manifold/cylinder O-ring during

reassembly using threaded guide rods to align the mating surfaces.

Description

On April 12, 2014, Unit 1 was in Mode 1 ascending in power after the 1R18 refueling outage. At approximately 20:08, control room operators received an MSIV

actuator trouble alarm followed by the MSIV not fully open indication. Control room

operators identified lowering loop 1 steam generator (SG) #1 level and steam flow and

manually tripped Unit 1 at about 28 percent reactor power. Upon further investigation, operators discovered a severe leak on the loop 1 outboard MSIV hydraulic actuator, which had caused the valve to close. Operators stabilized the plant in Mode 3 and all

safety related equipment responded as expected. The licensee assembled an issue

response team (IRT) and a root cause team to investigate the cause of the hydraulic

leak and subsequent manual reactor trip and to determine the required corrective

actions. Further investigation revealed that the manifold to cylinder O-ring on the valve

actuator had failed catastrophically due to being pinched during actuator reassembly in

2012. Further research by the root cause team revealed that maintenance personnel

relied on "skill of the craft" to install the O-ring and used a hoist to align the cylinder with

the manifold body. Use of the hoist resulted in rotational and/or oscillatory movement of

the mating surfaces, pinching the O-ring. The maintenance procedure that the

mechanics used to reassemble the actuator did not contain adequate instructions for

installing the manifold/cylinder O-ring. Specifically, maintenance procedure 26854-C, "Main Steam Isolation Valve Actuator Maintenance," which is used to perform

maintenance on Rockwell MSIVs, did not provide adequate instructions for installing the

lower manifold/cylinder O-ring during reassembly using threaded guide rods to align the

mating surfaces. The licensee revised the maintenance procedure, replaced the O-ring, and conducted an extent of condition evaluation of all other MSIV actuators. The

licensee entered this issue into their corrective action program as CR 800018.

Analysis:

The failure to provide adequate procedures required by 10 CFR 50 Appendix B Criterion V was a performance deficiency. The inspectors determined that the

performance deficiency was more than minor because it was associated with the

procedure quality attribute of the initiating events cornerstone and it adversely affected

the cornerstone objective to limit the likelihood of events that upset plant stability and

challenge critical safety functions during shutdown as well as power operations.

Specifically, the failure to provide adequate instructions for the installation of the

manifold to cylinder O-ring resulted in failure of the loop 1 MSIV and a subsequent

manual reactor trip due to lowering SG water level and steam flow. The inspectors

evaluated the finding using IMC 0609, Appendix A, "The Significance Determination

Process (SDP) for Findings At-Power," dated June 19, 2012. Because the inspectors

answered "No" to all the Exhibit 1, Section B, "Initiating Events Screening Questions,"

the inspectors determined that the finding was of very low safety significance (Green).

The inspectors determined the finding had a cross-cutting aspect of "resources" in the

human performance area, because the maintenance procedure used to install

manifold/cylinder O-ring did not provide adequate instructions for the proper installation

of the O-ring. [H.1]

Enforcement

10 CFR 50 Appendix B Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that procedures shall include appropriate quantitative or

qualitative acceptance criteria for determining that important activities have been

satisfactorily accomplished. Contrary to the above, the maintenance procedure used to

reassemble the MSIV hydraulic actuator did not provide adequate instructions for the

proper alignment of the manifold to cylinder mating surfaces resulting in a pinched O-

ring and subsequent MSIV actuator failure. Specifically, maintenance procedure 26854-

C, "Main Steam Isolation Valve Actuator Maintenance," which is used to perform

maintenance on Rockwell MSIVs, did not provide adequate instructions for installing the

lower manifold/cylinder O-ring during reassembly. To restore compliance, the licensee

revised the maintenance procedure, replaced the O-ring, and conducted an extent of

condition evaluation of all other MSIV actuators. This violation is being treated as an

NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. The violation was

entered into the licensee's corrective action program as CR 800018. (NCV

05000424/2014003-01, "Inadequate Maintenance Procedure Results in a Failed MSIV

and a Manual Reactor Trip")

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the five maintenance activities listed below to verify that 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 licensee's 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 licensee's planning and control of emergent work activities.

Documents reviewed are listed in the Attachment.

• Unit 2, week of May 5, 2014, Yellow risk condition associated with the extended allowed outage time (AOT) of the Unit 2 "A" EDG

• Unit 2, week of May 12, 2014, Orange risk condition associated with the extended AOT of the Unit 2 "A" EDG

• Unit 1, week of May 19, Yellow risk condition associated with the extended AOT of the Unit 1 "A" NSCW cooling tower fan #3

• Unit 1, week of June 2, 2014, during a planned maintenance outage of "1A" CCW pump in conjunction with an unplanned inoperability of the Unit 1A control room

emergency fan system (CREFS)

• Unit 2, week of June 16, 2014, Yellow risk condition associated with the extended AOT of the Unit 2 "B" EDG

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors selected the five 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 licensee's 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 776584, Unknown chemical buildup on top of upper motor windings

• CR 808990, "2B" EDG jacket water leak

• CR 805473/CAR 210188, 1HV3036A MSIV control board red light flickering

• CR 807567/CAR 210214, Unit 2 turbine driven auxiliary feedwater pump (TDAFW)controller output reading low

• CR 607966, U1 CCW Pump "1A" inboard bearing over 160 degrees Fahrenheit

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.

• SNC417397, Temporary modification to install accelerometers and a pressure transducer on chemical volume control system (CVCS) letdown lines, Unit 1

• DCP 98-VAN0055, Replace alternate radwaste building (ARB) with radwaste processing facility (RPF)

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 six 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.

• Maintenance Work Order (MWO) SNC137725 - Replacement of "1E" D26 relays MCC21805S3ABE

• MWO SNC413540 - 2PV3020 Replace A/B solenoid

• MWOs SNC408041 - (1A NSCW Fan 2) - Replace agastat relay and SNC383989 -

(1A NSCW Fan 2) - Replace rubber bushings on fan couplings

• MWO SNC525486 - Unit "2A" EDG Undervoltage relay calibration

• MWO SNC516991 - Unit 1 delta T/Tavg loop 3 protection channel operational test and calibration

• MWO SNC488414 - Unit 2 delta T/Tavg loop 1 protection channel I 2T-411 operational test and calibration 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.

• 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.

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

For the Unit 1 refueling outage which continued from April 1 2014, through April 27 2014, the inspectors evaluated the following outage activities:

• outage planning

• 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 The inspectors verified that the licensee:

• considered risk in developing the outage schedule

• controlled plant configuration in accordance with administrative risk reduction methodologies

• developed work schedules to manage fatigue

• developed mitigation strategies for loss of key safety functions

• adhered to operating license and technical specification requirements

Inspectors verified that safety-related and risk-significant structures, systems, and

components not accessible during power operations were maintained in an operable

condition. The inspectors also 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 seven surveillance tests 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

• 14802A-2 Rev. 5, Train "A" NSCW Pump / Check Valve IST and Response Time Test

• 24568-2 Rev. 38, RCP 1 Train "A", Reactor Trip Relays Under Frequency (281-A), Under Voltage (227-A), Timing (262R-A) Trip Actuating Device Operational Test and

Channel Calibration and 24565-2, Rev. 37, RCP 2 Train "A", Reactor Trip Relays

Under Frequency (281-A), Under Voltage (227-A), Timing (262R-A) Trip Actuating

Device Operational Test and Channel Calibration

• 24449-2 Rev. 9, Diesel Generator Power Out Train 2Q-2791 Channel Calibration

• 21118-2 Rev. 3.2, Centrifugal Charging Pump (CCP) Train "A" Safety Grade Charging Flow Loop 2F-0138 Channel Calibration

Reactor Coolant System Leak Detection

• 14905-1 Rev. 69.0, RCS Leakage Calculation (Inventory Balance)

• 14905-2 Rev. 53.0, RCS Leakage Calculation (Inventory Balance)

In-Service Tests (IST)

• 14804B-1 Rev. 5.0, Safety Injection Pump "B" Inservice and Response Time Tests

b. Findings

No findings were identified.

1EP6 Drill Evaluation

a. Inspection Scope

The inspectors observed the emergency preparedness drill conducted on May 21, 2014.

The inspectors observed licensee activities in the simulator and alternate technical

support center to evaluate implementation of the emergency plan, including event

classification, notification, and protective action recommendations. The inspectors

evaluated the licensee's performance against criteria established in the licensee's

procedures. Additionally, the inspectors attended the post-exercise critique to assess

the licensee's effectiveness in identifying emergency preparedness weaknesses and

verified the identified weaknesses were entered in the corrective action program.

b. Findings

No findings were identified.

2RS1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

Hazard Assessment and Instructions to workers During facility tours, the inspectors directly observed labeling of radioactive material and postings for radiation areas, HRAs

and airborne radioactivity areas established within the radiologically controlled area (RCA) of the Unit 1 containment, Unit 1 and Unit 2 auxiliary buildings, radwaste

processing facility, independent spent fuel storage installation, and selected storage

locations. The inspectors independently measured radiation dose rates or directly

observed conduct of licensee radiation surveys for selected RCA areas. The inspectors

reviewed survey records for several plant ar eas including surveys for alpha emitters, hot particles, airborne radioactivity, gamma surveys with a range of dose rate gradients, and

pre-job surveys for upcoming tasks. The inspectors also discussed changes to plant

operations that could contribute to changing radiological conditions since the last

inspection. For selected outage jobs, the inspectors attended pre-job briefings and

reviewed radiation work permit (RWP) details to assess communication of radiological

control requirements and current radiological conditions to workers.

Hazard Control and Work Practices The inspectors evaluated access barrier effectiveness for selected Unit 1 and Unit 2 locked high radiation area (LHRA) and very

high radiation area (VHRA) locations. Changes to procedural guidance for LHRA and

VHRA controls were discussed with HP supervisors. Controls and their implementation

for storage of irradiated material within the spent fuel pool were reviewed and discussed

in detail. Established radiological controls (including airborne controls) were evaluated

for selected Unit 1 refueling outage 18 (1R18) tasks including detensioning of the reactor

head, reactor head lift, upper internals lift, and scaffold building in Unit 1 containment. In addition, licensee controls for areas where dose rates could change significantly as a result of plant shutdown and refueling operations were reviewed and discussed.

Occupational workers' adherence to selected RWPs and HP technician (HPT)

proficiency in providing job coverage were evaluated through direct observations and

interviews with licensee staff. Electronic dosimeter (ED) alarm set points and worker

stay times were evaluated against area radiation survey results for detensioning of the

reactor head, reactor head lift, upper internals lift, and scaffold building in Unit 1

containment. ED alarm logs were reviewed and worker response to dose and dose rate alarms during selected work activities was evaluated. For HRA tasks involving

significant dose rate gradients, the inspectors evaluated the use and placement of whole

body and extremity dosimetry to monitor worker exposure.

Control of Radioactive Material The inspectors observed surveys of material and personnel being released from the RCA using small article monitor (SAM), personnel

contamination monitor (PCM), and portal moni tor (PM) instruments. The inspectors reviewed selected calibration records for selected release point survey instruments and

discussed equipment sensitivity, alarm setpoints, and release program guidance with

licensee staff. The inspectors compared recent 10 CFR Part 61 results for the dry active

waste (DAW) radioactive waste stream with radionuclides used in calibration sources to

evaluate the appropriateness and accuracy of release survey instrumentation. The inspectors also reviewed records of leak tests on selected sealed sources and

discussed nationally tracked source transactions with licensee staff.

Problem Identification and Resolution CRs associated with radiological hazard assessment and control were reviewed and assessed. The inspectors evaluated the

licensee's ability to identify and resolve the issues in accordance with procedure NMP-

GM-002, "Corrective Action Program," Version (Ver.) 12.1. The inspectors also

evaluated the scope of the licensee's internal audit program and reviewed recent assessment results.

Radiation protection activities were evaluated against the requirements of Updated Final

Safety Analysis Report (UFSAR) Section 12; TS Sections 5.4 and 5.7; 10 CFR Parts 19

and 20; and approved licensee procedures. Licensee programs for monitoring materials

and personnel released from the RCA were evaluated against 10 CFR Part 20 and IE

Circular 81-07, "Control of Radioactively Contaminated Material." Documents reviewed

are listed in the report Attachment.

b. Findings

Introduction

A Green, self-revealing, NCV of TS 5.7.1, "High Radiation Area," was identified for permitting an individual entry into a HRA without meeting the entry

requirements as specified in TS 5.7.1.b.

Description

On March 17, 2014, with the Unit 1 reactor shutdown for refueling, an operator was performing troubleshooting of Unit 1 containment sumps leakage. A

planned reactor coolant system crud burst was in progress on Unit 1. As a result of the

crud burst radiation levels in parts of the auxiliary building were elevated and areas were posted as high radiation areas. The operator observed a "Danger High Radiation Area -

HP Escort Required for Entry - Alarming Dosimetry," posting at the entrance to the

encapsulation vessel room and returned to the HP control point for further instructions.

The operator was briefed by an HP technician using a survey performed for the area on

March 6, 2014, that did not reflect the current postings or current radiological conditions.

The operator was informed by the HP technician that he could enter the area without an

HP escort because he was using an alarming ED. In the follow-up investigation the HP

technician stated that he was not aware the crud burst had started. Upon entry into the

encapsulation vessel room, the operator received a dose rate alarm on his ED. He

stopped immediately and exited the area. The worker's ED alarm setpoint was 250

millirem per hour (mrem/hr) and the highest exposure rate seen by the ED was 262 mrem/hr. Dose rates in the area were as high as 300 mrem/hr on contact and 193

mrem/hr at 30 cm based on a follow-up survey. The licensee entered this issue into

their corrective action program as CR 787908 and took immediate corrective actions

which included securing access to the area, performing follow-up surveys and convening

a human performance review board to examine causal factors for the purpose of

determining corrective actions.

Analysis:

The inspectors determined that entry into a HRA without meeting the entry requirements specified in T.S. 5.7.1 was a performance deficiency. This finding was

more than minor because it was associated with the occupational radiation safety

cornerstone attribute of human performanc e and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to

radiation from radioactive material during routine civilian nuclear reactor operation.

Specifically, workers permitted entry into HRAs with inadequate knowledge of current

radiological conditions could receive unintended occupational exposures. The finding

was evaluated using IMC 0609, Appendix C, "Occupational Radiation Safety

Significance Determination Process (SDP)", dated August 19, 2008. The finding was not

related to ALARA planning, nor did it involve an overexposure or substantial potential for

overexposure, and the ability to assess dose was not compromised. Therefore, the

inspectors determined the finding to be of very low safety significance (Green). The

inspectors noted that the operator responded properly to the ED dose rate alarm thereby

limiting his potential for unintended exposure. This finding had a cross-cutting aspect of

"avoid complacency" in the human performance area because HP personnel failed to

verify plant conditions through available m eans when an evolution was in progress that was known to increase area dose rates prior to authorizing entry into an HRA. [H.12]

Enforcement

Technical Specification 5.7.1, "High Radiation Area", requires in part, individuals entering HRAs meet one or more of the following criteria: a) be provided with

a radiation monitoring device that continuously indicates radiation dose rate in the area;

b) a radiation monitoring device that continuously integrates the radiation dose rate in

the area and alarms when a preset integrated dose is received. Entry into such areas

with this monitoring device may be made after the dose rate levels in the area have been

established and personnel are aware of them or c) An individual qualified in radiation

protection procedures with a radiation dose rate monitoring device, who is responsible

for providing positive control over the activities within the area and shall perform periodic

radiation surveillance at the frequency specified by health physics supervision in the

RWP. Contrary to the above, on March 17, 2014, a worker entered a HRA without a device that continuously indicated dose rates in the area (survey meter), knowledge of the actual radiological conditions in the area and no trained escort with a survey meter.

This violation is being treated as an NCV, consistent with Section 2.3.2 of the

Enforcement Policy. The violation was entered into the licensee's corrective action

program as CR 787908. (NCV 05000424, 2014003-02, "Unauthorized Entry into a High

Radiation Area.")

2RS2 Occupational ALARA Planning and Controls

a. Inspection Scope

Work Planning and Exposure Tracking The inspectors reviewed planned work activities and their collective exposure estimates for the current 1R18 outage. The inspectors

reviewed ALARA planning packages for the following high collective exposure tasks:

install/remove scaffolding, thermocouple work, mechanical stress improvement project (MSIP) work in containment and interference removal work in Unit 1 containment

annulus. For the selected tasks, the inspectors reviewed established dose goals, discussed assumptions regarding the bases for the current estimates with responsible

ALARA planners and walked down a mock-up of the reactor cavity annulus. The

inspectors evaluated the incorporation of exposure reduction initiatives and operating

experience, including historical post-job reviews, into RWP requirements. Day-to-day

collective dose data for the selected tasks were compared with established dose

estimates and evaluated against procedural criteria (work-in-progress review limits) for

additional ALARA review. Where applicable, the inspectors discussed changes to

established estimates with ALARA planners and evaluated them against work scope

changes or unanticipated elevated dose rates.

Source Term Reduction and Control The inspectors reviewed the collective exposure three-year rolling average from 2010 - 2012 and reviewed historical collective exposure trends from 1988 - 2014. The inspectors evaluated historical dose rate trends for

reactor coolant system piping and compared them to current 1R18 data. The crud burst

evolution during the first week of the 1R18 outage and source term reduction initiatives

were reviewed and discussed with chemistry and HP staff.

Radiation Worker Performance The inspectors observed radiation worker performance for job evolutions such as the MSIP interference removal, installation of shielding and

work in and around the reactor cavity. The inspectors observed ALARA briefings for

multiple MSIP jobs and emerging jobs such as Unit 1 bullet nose repair and radiation

worker performance was also evaluated as part of IP 71124.01. While observing job

tasks, the inspectors evaluated the use of remote technologies to reduce dose including

teledosimetry and remote visual monitoring.

Problem Identification and Resolution The inspectors reviewed and discussed selected corrective action program documents associat ed with ALARA program implementation.

The inspectors evaluated the licensee's ability to identify and resolve the issues in

accordance with licensee procedure NMP-GM-002, "Corrective Action Program", Ver.

12.1. The inspectors also evaluated the scope and frequency of the licensee's self-

assessment program and reviewed recent assessment results. ALARA program activities were evaluated against the requirements of UFSAR Section 12, TS Section 5.4, 10 CFR Part 20, and approved licensee procedures. Records reviewed are listed in

the report Attachment.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

a. Inspection Scope

Engineering Controls The inspectors reviewed the use of temporary and permanent engineering controls to mitigate airborne radioactivity during the 1R18 refueling outage.

The inspectors observed the use of portable air filtration units for work in contaminated

areas of the containment building and reviewed filtration unit testing records. The

inspectors evaluated the effectiveness of continuous air monitors and air samplers placed

in work area "breathing zones" to provide indication of increasing airborne levels.

Respiratory Protection Equipment The inspectors reviewed the use of respiratory protection devices to limit the intake of radioactive material. This included review of

devices used for routine tasks and devices stored for use in emergency situations. The

inspectors reviewed ALARA evaluations for the use of respiratory protection devices during

work associated with steam generator (S/G) eddy current testing. Selected self-contained

breathing apparatus (SCBA) units and negative pressure respirators (NPR)s staged for

routine and emergency use in the main control room and other locations were inspected for

material condition, SCBA bottle air pressure, number of units, and number of spare masks

and air bottles available. The inspectors reviewed maintenance records for selected SCBA

units for the past two years and evaluated SCBA and NPR compliance with National

Institute for Occupational Safety and Health certification requirements. The inspectors also

reviewed records of air quality testing for supplied-air devices and SCBA bottles.

The inspectors observed the use of powered air-purifying hoods during work on the S/G

platforms and in the upper cavity. The inspectors discussed training for various types of

respiratory protection devices with HP staff and interviewed radworkers and control room

operators on use of the devices. The inspectors reviewed respirator qualification records (including medical qualifications) for several main control room operators and emergency

responder personnel in the maintenance department.

Problem Identification and Resolution The inspectors reviewed CRs associated with airborne radioactivity mitigation and respiratory protection. The inspectors evaluated the

licensee's ability to identify and resolve the issues in accordance with licensee procedures.

The inspectors also reviewed recent self-assessment results.

Licensee activities associated with the use of engineering controls and respiratory

protection equipment were reviewed against TS Section 5.4; 10 CFR Part 20; Regulatory

Guide 8.15, "Acceptable Programs for Respiratory Protection," and applicable licensee

procedures. Documents reviewed are listed in the report Attachment.

b. Findings

No findings were identified.

2RS4 Occupational Dose Assessment

a. Inspection Scope

External Dosimetry The inspectors reviewed the licensee's national voluntary accreditation program (NVLAP) certification data for accreditation for the current year for

ionizing radiation dosimetry. The inspectors reviewed program procedures for

processing EDs and onsite storage of optically stimulated luminescent dosimeters (OSLD)s. Comparisons between ED and OSLD results, including correction factors, were discussed in detail. The inspectors also reviewed dosimetry occurrence reports

regarding alarming dosimeters.

Internal Dosimetry Inspectors reviewed and discussed the in vivo bioassay program with the licensee. Inspectors reviewed procedures that addressed methods for

determining internal or external contamination, releasing contaminated individuals, the

assignment of dose, and the frequency of measurements depending on the nuclides.

Inspectors reviewed and evaluated a sample of whole body counter (WBC) records

selected from September 2012 through February 2014. There were no internal dose

assessments for internal exposure greater than 10 millirem committed effective dose equivalent to review.

The inspectors evaluated the licensee's program for in vitro monitoring, however, no dose assessments had been performed using this method since the last inspection.

Special Dosimetric Situations The inspectors reviewed records for declared pregnant workers (DPW)s from September 2012 through February 2014 and discussed guidance

for monitoring and instructing DPWs. Inspectors reviewed and witnessed the licensee's

practices for monitoring external dose in areas of expected dose rate gradients, including the use of multi-badging and extremity dosimetry. The inspectors evaluated

the licensee's neutron dosimetry program incl uding instrumentation which was evaluated under procedure 71124.05. In addition, the inspectors evaluated the adequacy of

procedures and processes for assessing shallow dose.

Problem Identification and Resolution The inspectors reviewed and discussed licensee corrective action program documents associated with occupational dose assessment.

Inspectors evaluated the licensee's ability to identify and resolve the identified issues in

accordance with procedure NPM-GM-002, "Corrective Action Program", Ver. 12.1. The

inspectors also discussed the scope of the licensee's internal audit program and

reviewed recent assessment results.

Health physics program occupational dose assessment activities were evaluated against

the requirements of UFSAR Section 12; TS Section 5.4; 10 CFR Parts 19 and 20; and

approved licensee procedures. Records reviewed are listed in Section 2RS01, 2RS02, and 2RS04 of the report Attachment.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation

a. Inspection Scope

Radiation Monitoring Instrumentation

During walk-downs of the auxiliary building, radwaste processing building, fuel handling building and the RCA exit points, the

inspectors observed installed and portable radiation detection equipment. These

included area radiation monitors (ARM)s, cont inuous air monitors (CAMs), PCMs, SAMs, PMs, and liquid and gaseous effluent monitors, a WBC, count room equipment, and

portable survey instruments. The inspecto rs observed the physical location of the components, noted the material condition, noted flow measurement devices, input and

output of flow to monitors and compared sensitivity ranges with UFSAR requirements.

In addition to equipment walkdowns, the inspectors observed source checks and alarm

setpoint testing of various portable and fixed detection instruments including ion

chambers, a telepole, GEM TM-5s, ARGOS TM-ABs, and SAMs. Material condition of source check devices, device operation, and establishment of source check acceptance

ranges were also discussed with calibration lab personnel.

Calibration and Testing

The inspectors reviewed the last two calibration records for selected ARMs, PCMs, PMs, SAMs, and containment high-range ARMs and the most

recent calibration record for a WBC. Inspectors reviewed records of survey instrument

function/source checks and observed and discussed performance of required checks

with calibration lab personnel. Calibration source documentation was reviewed for the

ARM high-range calibrator and the Cs-137 (J.L. Shepherd) source used for portable instrument checks. Calibration stickers on portable survey instruments were reviewed

and inspections of storage areas for 'ready-to-use' equipment were completed during

walkdowns. The inspectors reviewed alarm se tpoint values for selected ARMs, PCMs, PMs, SAMs, and effluent monitors. The inspectors also reviewed count room quality

control records for germanium detectors and liquid scintillator counters.

Problem Identification and Resolution:

The inspectors reviewed selected CAP reports in the area of radiological instrumentation. The inspectors evaluated the licensee's ability

to identify and resolve the issues in accordance with procedure NMP-GM-002-001, "Corrective Action Program Instructions", Ver. 31.1.

Effectiveness and reliability of selected radiation detection instruments were reviewed

against details documented in the following: 10 CFR Part 20; NUREG-0737, "Clarification of TMI Action Plan Requirements"; UFSAR Chapters 11 and 12; and

applicable licensee procedures. Documents reviewed during the inspection are listed in

the report Attachment.

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 Unit 1 and Unit 2 PIs listed below. The inspectors reviewed plant

records compiled between April 2013 and March 2014 to verify the accuracy and

completeness of the data reported for the station. The inspectors verified that the PI

data complied with guidance contained in Nuclear Energy Institute 99-02, "Regulatory

Assessment Performance Indicator Guideline," and licensee procedures. The inspectors

verified the accuracy of reported data that were used to calculate the value of each PI.

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: Barrier Integrity

• reactor coolant system leak rate

• reactor coolant system specific activity

Cornerstone: Occupational Radiation Safety

The inspectors reviewed the occupational exposure control effectiveness PI results for

the occupational radiation safety cornerstone from January 2013 through December

2013. For the assessment period, the inspectors reviewed ED alarm logs and CRs

related to controls for exposure significant areas. Documents reviewed are listed in the

report Attachment.

Cornerstone: Public Radiation Safety:

The inspectors reviewed the radiological control effluent release occurrences PI results

for the public radiation safety cornerstone from January 2013 through December 2013.

The inspectors reviewed cumulative and projected doses to the public contained in liquid

and gaseous release permits and CRs related to radiological effluent technical

specifications/offsite dose calculation manual issues. The inspectors also reviewed

licensee procedural guidance for collecting and documenting PI data. Documents

reviewed are listed in the report Attachment.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review

The inspectors screened items entered into the licensee's corrective action program in

order to identify repetitive equipment failures or specific human performance issues for

follow-up. The inspectors reviewed condition reports, attended screening meetings, or

accessed the licensee's computerized corrective action database.

.2 Semi-Annual Trend Review

a. Inspection Scope

The inspectors reviewed issues entered in the licensee's corrective action program and

associated documents to identify trends that could indicate the existence of a more

significant safety issue. The inspectors focused their review on equipment issues, but

also considered the results of inspector daily condition report screenings, licensee

trending efforts, and licensee human performance results. The review nominally

considered the 6-month period of January 2014 through June 2014 although some

examples extended beyond those dates when the scope of the trend warranted. The

inspectors compared their results with the licensee's analysis of trends. Additionally, the

inspectors reviewed the adequacy of corrective actions associated with a sample of the

issues identified in the licensee's trend reports. The inspectors also reviewed corrective

action documents that were processed by the licensee to identify potential adverse

trends in the condition of structures, sy stems, and/or components as evidenced by acceptance of long-standing non-conforming or degraded conditions. Documents

reviewed are listed in the Attachment.

b. Findings and Observations

No findings were identified.

.3 Annual Follow-up of Selected Samples

a. Inspection Scope

The inspectors conducted a detailed review of condition report CR 648248, "Calculation Error Affects Emergency Action Level (EAL) Setpoints."

The inspectors evaluated the following attributes of the licensee's 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.

b. Findings

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 RG1 and Site Area Emergency Action

Level RS1 contained declaration threshold values which were significantly higher than

appropriate due to a calculation error.

Description:

In March 2005 Southern Co. corporate engineering calculation, X6CNA14, V3.0, 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 (Sieverts/second to mrem/hour) was made incorrectly and not detected during the

review process. The error resulted in threshold values sixty times greater than

appropriate. In 2005, Vogtle Electric Generating Plant submitted a license amendment

request to the NRC to change their EAL scheme to one based on NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," 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 29 of Vogtle procedure 91001-C, "Emergency Classification and Implementing Instructions," on March 20, 2008. The non-

conservative threshold values were contained in this implementing procedure.

During an extent of condition review prompted by Southern Co. fleet operating experience, calculation X6CNA14, V3.0 was found to contain the calculation error. On

May 31, 2013, this issue was placed in the licensee's corrective action program as CR

648248. The licensee took immediate corrective actions, which included providing

corrected threshold values to appropriate management and decision-makers (shift

managers/emergency directors). In addition, the licensee performed an enhanced

apparent cause determination per the licensee's procedures, processed formal changes

to the station emergency plan and associated implementing procedures, and performed

additional 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 September 2013.

These discrepant threshold values degraded the licensee's 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 (e.g.

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 licensee's ability to declare a Site Area

Emergency 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 lost 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 licensee's corrective action program as CR 648248. 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 in use by the nuclear facility licensee, and state and local response calls for reliance on information by

facility licensees for determinations of mi nimum 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 March 2008 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 classificati ons 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 11.05, criteria for 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 05000424, -425/2014003-03, "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)

.4 Operator Work-Around Annual Review

a. Inspection Scope

The inspectors performed a detailed review of the licensee's operator work-around, operator burden, and control room deficiency lists for the station in effect on June 16, 2014 to verify that the licensee identified operator workarounds at an appropriate

threshold and entered them in the corrective action program. The inspectors verified

that the licensee identified the full extent of issues, performed appropriate evaluations, and planned appropriate corrective actions. The inspectors also reviewed compensatory

actions and their cumulative effects on plant operation. Documents reviewed are listed

in the Attachment.

b. Findings

No findings were identified.

4OA3 Event Follow-up

.1 (Closed) Licensee Event Report 05000425/2014-001-00:

Automatic Reactor Trip Due to Low Steam Generator Level

a. Inspection Scope

On April 08, 2014, with Unit 2 in Mode 1, 100 percent reactor power, at approximately

04:28, operators received unexpected annunciators, "Digital Feedwater Trouble Alarm"

for all four steam generators. Upon further investigation, operators noted loop 3 steam

generator water level was lowering at rapid rate. The operator at the controls (OATC)

took manual control of the loop 3 MFRV and attempted to raise water level. Water level

continued to decrease to the SG low-low level reactor trip setpoint and an automatic

reactor trip occurred as expected. The inspectors reviewed the licensee event report (LER), the associated condition report and root cause determination, and subsequent

action items for potential performance deficiencies and/or violations of regulatory

requirements. Additionally, discussions were held with operations, engineering and

licensing staff members to understand the details surrounding this issue. This condition

was documented in the licensee's corrective action program as CR 797929. This LER is

closed.

b. Findings

Introduction

A Green, self-revealing non-cited violation (NCV) of 10 CFR 50 Appendix B Criterion V, "Instructions, Procedures, and Drawings," was identified for failure to

provide adequate work instructions as well as failure to follow the maintenance

procedure used to install flexible and rigid conduit. Specifically, the work instructions did

not provide adequate instructions and/or precautions to properly slope conduit during

installation to prevent water intrusion into a valve positioner. The work instructions

referenced maintenance procedure 25008-C, "Flexible and Rigid Conduit Installation."

The maintenance procedure referenced Vogtle design specification X3AR01 Section E-

8, "Raceway Systems," which provided sloping and tightness criteria for conduit installations.

Description

On April 08, 2014, with Unit 2 in Mode 1, 100 percent reactor power, at approximately 04:28, operators received unex pected annunciators, "Digital Feedwater Trouble Alarm," for all four steam generators. Upon further investigation, operators

noted loop 3 steam generator water level was lowering at rapid rate. The operator at the

controls (OATC) took manual control of the loop 3 MFRV and attempted to raise water

level. Water level continued to decrease to the SG low-low level reactor trip setpoint and

an automatic reactor trip occurred as expected. The plant was stabilized in Mode 3 and

all safety related equipment responded as expected. Loop 3 SG water level was

restored using auxiliary feedwater. The licensee subsequently assembled an issue

response team (IRT) and a root cause team to investigate the cause of the automatic

reactor trip due to the failure of the loop 3 MFRV and to determine the required

corrective actions. Further investigation revealed water had entered the loop 3 MFRV

positioner junction box through a conduit penetration from a leaking valve located

approximately twenty feet above the junction box. The water had shorted out the valve positioner and caused the MFRV to go shut. The licensee had identified the leak one month before the incident and had entered it into their corrective action program, but had

not yet entered it into the work control process. The licensee determined the conduit connection was loose and not installed per design specification drawing AX2D94V077-3, "Digital Feedwater Flow Controller Instrument Support Details," Rev. 1.0. The

specification drawing shows the conduit being routed to the underside of the junction

versus the top where it was installed. A combination of the loose conduit connection

combined with improper conduit installation resulted in the leaking water entering the

positioner junction box shorting the MFRV positioner and causing the MFRV to close.

Further research by the root cause team revealed that during digital feedwater design

modification installation, the work instructions used by the maintenance technician to

install the flexible conduit was inadequate. Specifically, the work instructions did not

contain sufficient detail to properly slope the conduit to prevent water intrusion. The

work instructions referenced maintenance procedure 25008-C, "Flexible and Rigid

Conduit Installation." The maintenance procedure directed the use of specification X3AR01 Section E-8, "Raceway Systems," which contained proper sloping and tightness

criteria. The licensee replaced the positioner, revised the procedure, and rerouted the

conduit per design specification. The licensee entered this issue into their corrective

action program as CR 797929.

Analysis:

The failure to provide adequate work instructions as well as the failure to follow maintenance procedure 25008-C as required by 10 CFR 50 Appendix B Criterion

V was a performance deficiency. The inspectors determined that the finding was more

than minor because it was associated with the procedure quality and human

performance attributes of the initiating ev ents cornerstone and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and

challenge critical safety functions during shutdown as well as power operations.

Specifically, the failure to provide adequate work instructions resulted in a failure of the

loop 3 MFRV and a subsequent automatic reactor trip due to low-low SG water level.

Using IMC 0609, Attachment 4, "Initial Characterization of Findings" dated June 19, 2012, the inspectors determined that the finding affected the initiating events

cornerstone. The inspectors evaluated the finding using IMC 0609, Appendix A, "The

Significance Determination Process (SDP) for Findings At-Power," dated June 19, 2012.

Because the inspectors answered "No" to all the Exhibit 1, Section B, "Initiating Events

Screening Questions," the inspectors determined that the finding was of very low safety

significance (Green). The inspectors determined that the finding had a cross-cutting

aspect of "procedure adherence" in the human performance area because the

maintenance electricians did not follow Vogtle design specification procedures or

drawings resulting in the improper sloping of the MFRV flexible conduit. [H.8]

Enforcement

10 CFR 50 Appendix B Criterion V requires, in part, that procedures shall include appropriate quantitative or qualitative acceptance criteria for determining that

important activities have been satisfactorily accomplished. Contrary to the above, the maintenance procedure used to install the flexible conduit during the Unit 2 digital

feedwater design change modification installation did not provide appropriate

instructions for the sloping and tightening of the conduit thus preventing water intrusion

into the loop 3 MFRV positioner junction box. Specifically, maintenance procedure

25008-C, "Flexible and Rigid Conduit Installation," which is used to install conduit, did not provide adequate instructions and/or precautions to properly slope and tighten conduit such that water intrusion is avoided. To restore compliance, the licensee

replaced the positioner, revised the procedure, and rerouted the conduit per the design

specification. This violation is being treated as an NCV, consistent with Section 2.3.2 of

the NRC Enforcement Policy. The violation was entered into the licensee's corrective

action program as CR 797929. (NCV 05000425/2014003-04, "Inadequate Maintenance

Procedures and Usage Results in a Failed MFRV and an Automatic Reactor Trip")

.2 (Closed) Licensee Event Report 05000424/2014-002-00:

Manual Reactor Trip Due to Main Steam Isolation Valve Failure

a. Inspection Scope

On April 12, 2014 Unit 1 was in Mode 1 ascending in power after the 1R18 refueling

outage. At approximately 20:08, control room operators received an MSIV actuator

trouble alarm followed by the MSIV not fully open indication. Control room operators

identified lowering loop 1 steam generator (SG) #1 level and steam flow and manually

tripped Unit 1 at about 28 percent reactor power. The inspectors reviewed the LER, the

associated condition report and root cause determination, and subsequent action items.

This condition was documented in the licensee's corrective action program as CR

800018. This LER is closed.

b. Findings

The enforcement aspects associated with this event are discussed in Section 1R12 of

this integrated inspection report.

4OA5 Other Activities

.1 (Closed) Unresolved Item 05000425/2013007-02:

Failure to Identify and Correct Potential Emergency Diesel Generator "2B" Inoperability Following Failed Surveillance Testing

a. Inspection Scope

During the component design bases inspection documented in NRC Inspection Report

05000424, 425/2013007 (ADAMS ML13269A419), the team identified an unresolved

item (URI) regarding the discovery of a condition that could have potentially resulted in

an inoperable condition of the "2B" EDG due to an intermittently misaligned mechanically

operated cell (MOC) switch. Since the licensee had not recognized the potential

operability impact on the "2B" EDG during their investigations of EDG surveillance test

failures on December 13, 2011, and June 25, 2012, additional NRC inspection of the

specific alignment of the affected MOC switch contacts, and of the licensee's evaluation

of operability of the "2B" EDG, prior to the MOC switch being adjusted, was necessary to

determine if the issue of concern was minor or more than minor. On June 16, 2014, NRC inspection of the MOC switch contacts was performed to determine if the proper

functioning of the "2B" EDG, during emergency mode of operation, would have been

affected. Based on this additional review, this URI is now closed.

b. Findings

No findings were identified. However, the inspectors identified a minor performance deficiency and associated minor violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." In accordance with IMC 0612, "Power Reactor Inspection Reports,"

dated January 24, 2013, minor violations are not routinely documented in inspection

reports. However, they may be documented to discuss inspection activities and

conclusions for closing a URI.

The inspectors determined that the licensee's failure to promptly identify and correct a

misaligned MOC switch associated with the "2B" EDG output breaker following a

surveillance test failure on December 13, 2011, was contrary to 10 CFR 50, Appendix B, Criterion XVI, and was a performance deficiency. This failure led to a small amount of

additional unavailability to troubleshoot the issue following an additional failure on June

25, 2012. Following additional NRC inspection on June 16, 2014, the inspectors

determined the actual radial alignment of the MOC switch contacts would have

supported the proper functioning of the EDG if it had been called upon during an event.

Using IMC 0612, Appendix B, "Issue Screening," dated September 7, 2012, the

inspectors determined the issue was of minor significance because, if left uncorrected, would not have led to a more significant safety concern. The licensee corrected the

condition of the misaligned MOC switch following the second failure on June 25, 2012.

Because this issue was entered into the licensee's corrective action program as CR

687752, and was of minor significance, the failure to comply with 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," constituted a minor violation that is not subject to

enforcement action in accordance with the NRC's Enforcement Policy.

4OA6 Meetings, Including Exit

.1 Exit Meeting

On July 25, 2014, the resident inspectors presented the inspection results to

Mr. T. Tynan and other members of the licensee's staff. The inspectors confirmed that

proprietary information was not provi ded or examined during the inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

R. Barringer, Security Manager

R. Collins, Chemistry Manager

H. Cooper, Engineering Programs Supervisor

J. Dixon, Corporate Fleet Area Manager, Health Physics

G. Gunn, Licensing Supervisor

M. Hayden, EP Manager

R. Hons, Training Manager

M. Johnson, Health Physics Manager

K. Morrow, Licensing

F. Pournia, Engineering Director

J. Robinson, Engineering Programs Manager

I. Sarygin, Sr. Engineer

G. Saxon, Plant Manager

J. Thomas, Work Management Director

T. Thompson, Systems Engineering Manager

T. Tynan, Site Vice-President

K. Walden, Licensing Engineer

S. Waldrup, Licensing Director

NRC personnel

F. Ehrhardt, Chief, Region II Reactor Projects Branch 2

LIST OF ITEMS

OPENED AND CLOSED

Opened

05000424,425/2014003-03 VIO Calculation Error Results in Significantly Non-

Conservative EAL Threshold Values (Section

4OA2.3)

Open and

Closed

05000424/FIN-2014003-01 NCV Inadequate Maintenance Procedure Results in a

Failed MSIV and a Manual Reactor Trip (Section

1R12)

05000424/FIN-2014003-02 NCV Unauthorized Entry into a High Radiation Area (Section 2RS1)

05000425/FIN-2014003-04 NCV Inadequate Maintenance Procedures and Usage

Results in a Failed MFRV and an Automatic

Reactor Trip (Section 4OA3.1)

Attachment

Closed

05000424/FIN-2014003-01 NCV Inadequate Maintenance Procedure Results in a

Failed MSIV and a Manual Reactor Trip (Section

1R12)

05000424/FIN-2014003-02 NCV Unauthorized Entry into a High Radiation Area (Section 2RS1)

05000425/FIN-2014003-04 NCV Inadequate Maintenance Procedures and Usage

Results in a Failed MFRV and an Automatic

Reactor Trip (Section 4OA3.1)

Attachment

Section 1R01: Adverse Weather Protection

Procedures

11889-C Rev. 21, Severe Weather Checklist

VNP-CMS-710-00-PR-00001 Rev. 0, CB&I Heal th, Safety and Environmental Management System (Units 3&4)

VNP-CMS-710-03-PR-00400 Rev. 0, CB&I Emergency Preparedness Plan (Units 3&4)

230-1, Rev. 23.0, Offsite AC Circuit Verification and Capacity/Capability Evaluation

230-2, Rev. 22.0, Offsite AC Circuit Verification and Capacity/Capability Evaluation

18017-C, Rev. 9.6, Abnormal Grid Disturbances/Loss of Grid

13830-1, Rev. 69.0, Main Generator Operation

13830-2, Rev. 55.0, Main Generator Operation

Section 1R04: Equipment Alignment

Procedures

11145-2 Rev. 12.2, Diesel Generator Alignment

11146-2 Rev. 7.1, Diesel Generator Fuel Oil Transfer System Alignment

11610-2 Rev. 21.3, Auxiliary Feedwater System Alignment

Drawings

2X4DB170-1 Rev. 42.0, P&I Diagram Diesel Generator System Train A - System No. 2403 2X4DB170-2 Rev. 47.0, P&I Diagram Diesel Generator System Train B - System No. 2403 2X4DB161-1, P&I Diagram Rev. 36.0, Aux iliary Feedwater System Condensate Storage &

Degasifier System, System No. 1302

2X4DB161-2, P&I Diagram Rev. 24.0, Au xiliary Feedwater Syst em, System No. 1302 2X4DB161-3, P&I Diagram Rev. 38.0, Auxilia ry Feedwater Pump System, (Aux Feedwater Pump Turbine Driver) System No. 1302

2X4DB168-3, P&I Diagram Rev. 37.0, C

ondensate and Feedwater System, System No.

1305

Section 1R05: Fire Protection

Procedures

2754-2 Rev. 0.2, Zone 54 - Auxiliary Building - Level 2 Train "A" CCW HX Fire Fighting

Preplan Attachment 92755-2 Rev. 0.2, Zone 55 - Auxiliary Building - Level 2 Train "B" CCW HX Fire Fighting Preplan

2848-2 Rev. 0.2, Zone 148 - Auxiliary Building - Level 2 Fire Fighting Preplan

2723-2 Rev. 2.1, Zone 23 - Auxiliary Building - Electrical Chase Rooms Fire Fighting Preplan

2872-2 Rev. 1.2, Zone 172 - Auxiliary Building - Level 2 Fire Fighting Preplan

2847-2 Rev. 1.2, Zone 147 - Auxiliary Building - Level 2 Fire Fighting Preplan

2714B-1 Rev. 2.2, Zone 14B - Auxiliary Building - Level C Fire Fighting Preplan

2719-1 Rev. 4.1, Zone 19 - Auxiliary Building - CVCS Centrifugal

Charging Pump Rooms

Fire Fighting Preplan

2720-1 Rev. 4.1, Zone 20 - Auxiliary Building - CVCS Pump Rm Train A Fire Fighting Preplan

2721-1 Rev. 5.1, Zone 21 - Auxiliary Building - CVCS NCP Room Fire Fighting Preplan

2789-2 Rev. 3.1, Zone 89 - Control Building - Level A Fire Fighting Preplan

2790-2 Rev. 2.2, Zone 90 - Control Building - Level A Fire Fighting Preplan

2859-2 Rev. 1.2, Zone 159 - Control Building - Level A Fire Fighting Preplan

2787-2 Rev. 2.2, Zone 87 - Control Building - Level A Fire Fighting Preplan

2788-2 Rev. 2.2, Zone 88 - Control Building - Level A Fire Fighting Preplan

2793-2 Rev. 3.2, Zone 93 - Control Building - Level A Fire Fighting Preplan

2802-2 Rev. 2.2, Zone 102 - Control Building - Level A Fire Fighting Preplan

2858-2 Rev. 1.2, Zone 158 - Control Building - Level A Fire Fighting Preplan

2862-1 Rev. 2.2, Zone 162 - Diesel Generator Building Fire Fighting Preplan

2864-1 Rev. 2.2, Zone 164 - Diesel Generator Building - Train B DFO Tank Fire Fighting

2855-2, Rev. 0.2, Zone 155 - Auxiliary Feedwater Pumphouse - Train B Fire Fighting Preplan

2856-2, Rev. 0.2, Zone 156 - Auxiliary Feedwater Pumphouse Fire Fighting Preplan

2857A-2, Rev. 0.2, Zone 157A - Auxiliary Feedwater Pumphouse - Train C Fire Fighting

Preplan

2857B-2, Rev. 0.2, Zone 157B - Auxiliary Feedwater Pumphouse - Train C Fire Fighting

Preplan

Section 1R06: Internal Flooding

Procedures

219-1 Rev. 35, Auxiliary and Containment Buildings and Miscellaneous Drain Systems

Other X6CXC-27 Rev.8, Flooding Analysis Auxiliary Building Level D

AX1D94A56 Rev. 2.0, Auxiliary Building Units 1 & 2 Door Schedule Level D

CCN-V-07-0011 Rev. 8.0, Flooding - Auxiliary Building Level D

Drawings

AX1D08A02-2, Rev. 6.0, Auxiliary Building Floor Plan El. 119 Level D

Section 1R11: Licensed Operator Requalification Program

Procedures

2003-C Rev. 53, Reactor Startup Mode 3 to Mode 2

2004-C Rev.107.2, Power Operation Mode 1

NMP-OS-007-001 Rev. 14.3, Conduct of Operations Standards and Expectations

Attachment Other Simulator scenario V-RQ-SE-12702, Loss of Grid/Natural Circulation Cooldown

Simulator scenario V-RQ-SE-14300, Pe rformance Improvement Exercise Simulator scenario V-RQ-SE-14301, Large Break LOCA Response

Simulator scenario V-RQ-SE-14302, SGTL/SGTR/Recovery

Simulator scenario V-RQ-SE-14303, Control Room Evacuation

Section 1R12: Maintenance Rule Effectiveness

Condition Reports

and Action Items

807906, MPFF documented for Unit 2, System 1305, 2HV5230

795933, Unexpected control room annunciator ALB16-D04, MFIV Loop 4 low hydraulic pressure

Section 1R15: Operability Evaluations

Condition Reports

776584, Unknown chemical buildup on top of upper motor windings

808990, 2B DG Jacket Water Leak

805473/CAR

210188, 1HV3036A MSIV control board 'red' light flickering

807567/CAR

210214, Unit 2 turbine driven auxiliary feedwater pump (TDAFW) controller output

reading low

CR 607966, U1 CCW Pump 1A inboard bearing over 160F

Other Records

TE 776816, IDO request for 1A RHR pump

TE 776799, "OBDN resolution" 1A RHR IDO comp action

TE 776807, "OBDN resolution" 1A RHR IDO comp action

EMI Diagnostics Report for Plant Vogtle 1 and 2 Electric Generating Plant, by Doble Global

Power Services, PO# SNG10075822 dated 3/10/2014 initiative

CAR 210245, IDO - 2B DG Jacket Water Leak

MWO SNC572497, 2B DG Jacket Water Leak

CAR 210188, IDO - 1HV3036A MSIV

CAR 210214, IDO - Unit 2 turbine driven auxiliary feedwater pump (TDAFW) controller

MWO SNC525698, Troubleshoot Unit 2 TDAFW controller output

TE 767342, IDO revision for CR 607966

Section 1R18: Plant Modifications

Procedures

NMP-AD-010 Rev. 13.0, 10

CFR 50.59 Screening/Evaluation

NMP-ES-054-001 Rev. 2.0, Temporary Modification Processing

Work Orders

SNC417397, Temporary modification to install accelerometers and a pressure transducer on

CVCS letdown lines, Unit 1

1081013501, Accelerometer Installation at the CVCS letdown flow orifices and line 1-1208-255-

3", 6/13/2008

DCP 98-VAN0055, Replace the Alternate Radwaste Building (ARB) with the Radwaste

Processing Facility (RPF)

Attachment

Drawings

AX3D-CH-T01J, Wiring Diagram Alternate Radwaste Building and ABB Control Room Misc Devices

AX3D-BC-G20C, Elementary Diagram Alternate R

adwaste Building Cabling Block Diagram Rad Monitors, HVAC, Bridge Crane

AX3DH469-1, Wiring Diagram Alternate Radwaste Building Control Room Conduit and Lighting

and Communications Plans Sheet 001

Corrective Action Documents

Condition Report (CR)

2008106194, Walk down of Unit 1 containment for increased leakage

discovered upstream of letdown orifice isolation valve 1HV8149A, 6/1/2008

Technical Evaluation (TE) 34658, Corrective Action to establish a replacement interval for the

letdown flow orifices, 5/7/2009

TE 14379, Corrective action to replace the Unit 1 letdown flow orifices with a butt weld

connection orifice during the refueling outage 15, 9/11/2008

Enhanced Apparent Cause Determination (EACD)

194554, Station personnel failed to

implement the corrective action program to resolve an uncontrolled change in which area radiation monitors were permanently removed fr om the Alternate Radwaste Building (ARB).

Technical Evaluation (TE)

363628, Revise procedure

NMP-GM-002-001 Attachment 1 to

provide guidance for screening CRs that include design document aspects and configuration

control issues.

TE 366691, Generate an LDCR to update the FSAR to reflect the

ARE-16851,

ARE-16852,

ARE-16853,

ARE-16854 as being no longer in service.

TE 366715, Complete and approve an ABN to update any associated documents to reflect the

ARE-16851,

ARE-16852,

ARE-16853,

ARE-16854 as being no longer in service.

TE 367763, Properly label as abandoned in place or remove all remnants of the ARB rad

monitor system that is no longer in use.

Other

VEGP-FSAR-11, Radioactive Waste Management

VEGP-FSAR-12, Radiation Protection

ABN-V03007, Incorporate PDMS changes per DEC

DBN-V03007

LDCR No.

2012017, Update the FSAR to reflect the

ARE-16851,

ARE-16852,

ARE-16853 and

ARE-16854 as being no longer is Service.

Section 1R19: Post Maintenance Testing

Procedures

14825-2 Rev. 94, Quarterly Inservice Valve Test

14825-2 Rev. 95, Quarterly Inservice Valve Test

14430-1 Rev. 11.0, NSCW Cooling Tower Fans Monthly Test

24449-2 Rev. 9, Diesel Generator Power Out Train 2Q-2791 Channel Calibration

24812-1 Rev. 44, Unit 1 Delta T/Tavg loop 3 protection channel III 1T 431 operational test and

calibration

24810-2 Rev. 36, Unit 2 Delta T/Tavg loop 1 protection channel I 2T-411 operational test and

calibration

Attachment

Work Orders

SNC137725 - Replacement of 1E D26 Relays MCC21805S3ABE

SNC413540 - 2PV3020 Replace A/B Solenoid

SNC527135 - Quarterly Steam Generator Atmos pheric Relief Valve Inservice Valve Test

SNC507135 - Manually stroke 2PV3020 from the loca

l control station and perform ARV fail safe test per 14825-2

SNC408041 - (1A NSCW Fan 2) - Replace Agastat Relay

SNC383989 - (1A NSCW Fan 2) - Replace Rubber Bushings on Fan Couplings

SNC525486 - Unit 2A EDG Undervoltage Relay Calibration

SNC516991 - Unit 1 Delta T/Tavg loop 3 protection channel operational test and calibration

SNC488414 - Unit 2 Delta T/Tavg loop 1 protection channel I 2T-411 operational test and

calibration

Other Records Unit 2 operator logs for 4/14/14

Unit 2 operator logs for 4/26/14

Unit 2

ARV 3020 system outage fragnet

Unit 1 operator logs for 5/12/14

1A NSCW Fan 2 system outage fragnet

Section 1R22: Surveillance Testing

Procedures

14802A-2 Rev. 5, Train A NSCW Pump / Check Valve IST and Response Time Test

24568-2 Rev. 38, RCP 1 Train A, Reactor Trip Relays Under Frequency (281-A), Under Voltage

(227-A), Timing (262R-A) Trip Actuating Device Operational Test and Channel Calibration

24565-2, Rev. 37, RCP 2 Train A, Reactor Trip Relays Under Frequency (281-A), Under

Voltage (227-A), Timing (262R-A) Trip Actuating Device Operational Test and Channel

Calibration

14804B-1 Rev. 5.0, Safety Injection Pump B Inservice and Response Time Tests

24449-2 Rev. 9, Diesel Generator Power Out Train 2Q-2791 Channel Calibration

21118-2 Rev. 3.2, Centrifugal Charging Pump (CCP) Train A Safety Grade Charging Flow Loop

2F-0138 Channel Calibration

14905-1 Rev. 69.0, RCS Leakage Calculation (Inventory Balance)

14905-2 Rev. 53.0, RCS Leakage Calculation (Inventory Balance)

Work Orders

SNC523019 - Quarterly train A NSCW pump 21202P4005 discharge MOV and check valve

inservice test

SNC523018 - Quarterly train A NSCW pump 21202P4003 discharge MOV and check valve

inservice test

SNC523447- Quarterly train A NSCW pump 21202P4001 discharge MOV and check valve

inservice test

SNC528899, Quarterly train A RCP #1 under voltage and under frequency relays TADOT

SNC405763, 18-month train A RCP #2 under voltage and under frequency relays TADOT

SNC457082, 18M staggered test basis (train B) safety injection pump response time test

SNC520727, Quarterly (train B) safety injecti on pump and discharge check valve inservice test

SNC525486, Unit 2A EDG Undervoltage Relay Calibration

SNC412442, Centrifugal Charging Pump (CCP) Train A Safety Grade Charging Flow Loop 2F-

Attachment

0138 Channel Calibration

Section 2RS1: Radiological Hazard Assessment and Exposure Controls

Procedures, Guidance Documents, and Manuals

00008-C, Plant Lock and Key Control, Ver. 16.2

11882-1, Outside Area Rounds Sheets, Ver. 90.2

43014-C, Special Radiological Controls, Ver. 43.5

43021-C, Health Physics Central Monitoring Station Expectation and Guidelines, Rev. 4.4

43022-C, Health Physics Central Monitoring Station, Ver. 5.2

43032-C, Reactor Head and Upper Internals Movement, Ver. 3.2

46100-C, 10

CFR 61 Waste Classification Sampling Program, Ver. 9

46111-C, Storage of Radwaste in Outdoor Process Shields, Ver. 6.1

47009-C, Operation and Use of Portable Ventilation Units, Ver. 22.3

93610-C, Conduct of Special Nuclear Material Control and Accountability, Ver. 11.1

93641-C, Development and Implementation of the Fuel Shuffle Sequence Plan, Ver, 21.1

93780-C,

HI-TRAC Contamination Survey, Ver. 1.0

93781-C, HI -TRAC Surface Dose Rates, Ver. 1.0

93782-C,

HI-STORM Surface Dose Rates, Ver. 1.0

NMP-GM-002, Corrective Action Program, Ver. 12.1

NMP-GM-002-001, Corrective Action Program Instructions, Ver. 31.1

NMP-HP-109, Investigation, Evaluation and Management of Damaged, Lost, Malfunctioning or

Alarming Dosimetry, Ver. 1.1

NMP-HP-202, Radiological Controls for Highly Radioactive Objects, Ver. 1.0

NMP-HP-206, Issuance, Use and Control of Radiation Work Permits, Ver. 3.0

NMP-HP-207, Selection and Use of Protective Clothing, Ver. 1.0

NMP-HP-218, Health Physics Stop Work Authority and Guidance on Response, Ver. 1.0

NMP-HP-300, Radiation and Contamination Surveys, Ver. 2.1

NMP-HP-301, Airborne Radioactivity Sampling and Evaluation, Ver. 2.2

NMP-HP-302, Restricted Area Classification, Postings, and Access Control, Ver. 6.0

NMP-HP-302-001, Radiological Key Control, Ver. 2.1

NMP-HP-303, Personnel Decontamination, Ver. 2.2

NMP-HP-304, Decontamination of Areas, Tools and Equipment, Ver. 1.0

NMP-HP-305, Alpha Radiation Monitoring, Ver. 4.0

NMP-HP-400, Control and Accountability of Radioactive Sources, Ver. 2.0

NMP-HP-403, Control and Monitoring of Materials in Radiation Controlled Areas, Ver. 1.0

NMP-HP-404, Release of Materials from the RCA and Protected Areas, Ver. 1.0

Health Physics Work Plan, Rx Cavity Decon

Records and Data

46100-C, 10

CFR 61 Waste Classification Sampling Program, Ver. 9, Dated 06/12/12

Air Sampler Calibration, Sheet 1 of 3, Data Sheet 1, Air Sampler Calibration Form, Instrument

Nos.

VEGP-HP-1368, Model No.

RAS-1, Dated 03/13/14;

VEGP-HP-1369, Model No.

RAS-1, Dated 01/09/14; and

VEGP-HP-1371, Model No.

RAS-1, Dated 12/26/13

Airborne Radioactivity Sampling and Evaluation, Data Sheet 1, Air Sample Record (Particulate and Iodine), Air Sample Nos.:

14-0132, U1 CTMT/220' (Pulling NI Covers in Upper RX Cavity), Dated 03/16/14; 14-0133, U1 EH (Equipment Hatch Routine), Dated 03/16/14; and 14-0157, U1

CTMT/171' (Routine

VEGP-HP-1368), Dated 03/20/14

National Source Tracking System, Annual Inventory Reconciliation Report, Vogtle 1, Dated

01/17/14

Attachment Plant Vogtle, Gamma Spectroscopy Results, Sample IDs:

86362, U1 RX Head Lift Level 220

(1L Gas sample in liquid marinelli), Dated 03/19/14; 86363, U1 RX Head Lift (1L Gas sample in liquid marinelli), Dated 03/19/14; 86375, U1 Polar Crane RX Head Lift (Particulate Shelf 0), Dated 03/19/14; 86376, U1 Polar Crane RX Head Lift (Breathing Zone Charcoal Shelf 0), Dated

03/19/14; 86406 and 86409, 1-CTMT 220'-South Cavity-Upper Internal Lift (Particulate Shelf 0), Dated 03/20/14; and 86407 and 86408, 1-CTMT 220'-South Cavity-Upper Internal Lift (Large

Plastic Charcoal Shelf 0), Dated 03/20/14

Plant Vogtle Radiological Information Survey Nos.

165158,

HI-TRAC Surface Dose (C), Dated

11/22/13;

165176, HI Storm Surface Dose Rates (C), Dated 11/22/13;

165177, HI Storm Duct

Survey C, Dated 11/22/13;

165641,

HI-TRAC Surface Dose (C), Dated 12/11/13;

165655, HI

Storm Duct Survey C, Dated 12/11/13;

168234, Upper Cavity (1RXA16), Dated 03/16/14;

168472, Reactor Cavity Area (1RXA2), Dated 03/19/14;

168526, Quadrant 3 (1RXC), Dated

03/20/14;

168528, Reactor Cavity Area (1RXA2), Dated 03/19/14;

168538, Reactor Cavity Area

(1RXA2), Dated 03/20/14;165662, ISFSI Pad (C), Dated 12/11/13,

169489, U1 Upper Cavity,

4/4/14,

169524, U1 Upper Cavity, 4/4/14 and

169517, U1 Upper Cavity, 4/4/14

RWP No. 14-1006, Installation and Removal of Insulation in Unit 1 Containment, Revision

(Rev.) 0

RWP No. 14-1403, Decon of Upper and Lower Cavity, Rev. 0

RWP No. 14-1406, Reactor Head and Upper Internals Lift and Set, Rev. 0

RWP 14-1612, MSIP Interference Removal and Support Activities in U1 Cnmt Annulus, Rev. 0

Unit 1 and U2 Spent Fuel Pool Inventory Log, Non Fuel Radioactive Material Stored in Unit 1

and U2 Spent Fuel Pool, Dated 02/18/14

Unitech Services Group, Customer Provided HEPA Filter Testing Maintenance Log, HEPA Unit

Type:

Portable, HEPA S/Ns:

HU2000, Dated 03/06/14; HU200002, Dated 03/06/14; and HU

35015, Dated 03/06/14

CAP Documents

CR 603893

CR 604563

CR 610824

CR 615028

CR 624795

CR 663674

CR 679060

CR 697578

CR 787908

CR 795074

Health Physics Fleet Performance Summary Report,

NOSCPA-HP-2013-13, Dated 12/04/13

Nuclear Oversight Audit of Health Physics, Fleet-HP-2013, Dated July 15, 2013

Section 2RS2: Occupational

ALARA Planning and Controls Procedures, Guidance Documents, and Manuals

16035-1, "Chemistry Operations Interface for RCS Chemistry Control During Scheduled Plant

Shutdowns", Ver. 15.2

NMP-AD-035, "ALARA Program", Ver. 1.3

NMP-HP-204, "ALARA Planning and Job Review", Ver. 3.3

Attachment

41006-C, Temporary Shielding, Ver 29.2

NMP-HP-202, Radiological Controls for Highly Radioactive Objects, Ver. 1.0

NMP-HP-206, Issuance, Use and Control of Radiation Work Permits, Ver 3.0

Records and Data

U-1 Containment 1R18 Outage Turnover, dated 03/20/2014

1R18 Outage Dose Summary Report, dated 03/20/2014

HP Duty Foreman's Checklist - Daily Report Items, dated 03/19/2014

1R18 Temporary Shielding Worksheet, dated 11/21/2013

Plant Vogtle Radiological Information Survey Nos.

168301, Under Vessel Annulus Area (Pre-

Shielding), Dated 03/17/2014,

168313, Under Vessel Annulus Area (Pre-Shielding), Dated

03/18/2014,

168345, Under Vessel Annulus Area (Post Shielding), Dated 03/18/2014,

168362, Under Vessel Annulus Area (Post Shielding), dated 03/17/2014,

169356, Reactor Cavity Area

for Core Exit Thermocouple Bullet Nose Manual Alignment, dated 04/02/2014,

169077, Lower

Reactor Cavity Area for Core Exit Thermocouple Bullet Nose Manual Alignment, dated

03/29/2014, Plant Vogtle EPRI Radiological Survey Nos.

168339, EPRI Survey Map Loop 1, dated

03/18/2014,

168477, EPRI Survey Map Loop 1, dated 03/20/2014,

168339, EPRI Survey Map Loop 2, dated 03/18/2014,

168488, EPRI Survey Map Loop 2, dated 03/20/2014,

168341, EPRI

Survey Map Loop 3, dated 03/18/2014,

168491, EPRI Survey Map Loop 3, dated 03/20/2014,

168337, EPRI Survey Map Loop 4, dated 03/18/

2014,

168478, EPRI Survey Map Loop 4, dated

03/20/2014,

Work in Progress (WIP) Reviews,

RWP 14-1004, Installation and Removal of Scaffolding in U1

Containment (50%), dated 03/23/2014,

RWP 14-1408, Theermocouple Work in U1

Containment, dated 03/28/2014,

RWP 14-1611, MSIP Westinghouse Squeeze (50%), dated

03/26/2014,

RWP 14-1611, MSIP Westinghouse Shim Gap Work (50%), dated 03/28/2014, ,

RWP 14-1612, MSIP Interference Work (50%), dated 03/25/2014,

RWP 14-1612, MSIP

Interference Removal (80%), dated 03/31/2014

ALARA Briefing Records,

RWP 14-1004, Install/Remove Scaffold in U1 CTMT,

RWP 14-1408, Thermcouple Work in Containment,

RWP 14-1611, MSIP Westinghouse, ,

RWP 14-1612, MSIP

Interference Work

ALARA Post Job Reviews,

RWP 13-2004, Install/Remove Scaffold in U2 CTMT,

RWP 13-2302, Eddy Current Testing on S/G 1&2 and All Associated Work,

RWP 13-2400, Rx Head Disassembly/Assembly

Shutdown Chemistry Review: Vogtle Unit 1 Fuel Cycle 17, dated 11/27/2012

U1 EPRI Shutdown Survey Points Trend Graph for Refueling Outages 1R1 - 1R17

U1 S/G Channel Head Dose Rate Trend Graph for Refueling Outages 1R1 - 1R17

U2 EPRI Shutdown Survey Points Trend Graph for Refueling Outages 2R1 - 2R16

U2 S/G Channel Head Dose Rate Trend Graph for Refueling Outages 2R1 - 2R16

VEGP ALARA Strategic Plan 2013 - 2018

EPRI Sponsored Source Term Assessment for Vogtle Units 1 and 2, Final Report, Dec 2013

NOSCPA-HP-2012-04, Health Physics Fleet Performance Summary Report, dated 11/26/2012

VNP - Health Physics Focused Self Assessment for Dose Controls, dated 01/02/2013

NOSCPA-HP-2013-13, Health Physics Fleet Performance Summary Report, dated 12/04/2013

ALARA Committee Meeting Minutes Fourth Quarter 2013

2012 Annual ALARA Report, 09/25/2013

1R17 ALARA Report Attachment

2R16 ALARA Report Plant ALARA Review Committee (PARC) "Called Monthly Meeting," dated 03/21/14

PARC "Called Monthly Meeting," dated 03/05/14

RWP Dose Totals Year to Date (YTD), dated 04/03/14

CAP Documents

CR 610495

CR 643120

CR 650993

CR 651612

CR 762528

CR 763764

Section 2RS3: In-Plant Airborne Radioactivity Control and Mitigation Adam 4/9/14

Procedures, Guidance Documents, and Manuals

47020-C, DOP Testing of HEPA Filters, Ver. 5.2

47004-C, Breathing Air Analysis, Rev. 16

47001-C, Selection and Use of Respiratory Protection Equipment Used for Radiological

Purposes, Ver. 19.2

47005-C, Inspection, Repair, and Storage of Respiratory Protection Equipment, Rev. 15

NMP-GM-002-001, Corrective Action Program, Ver. 31.1

Records and Data Reviewed

SCBA Maintenance Records, Kit 58 and

HP-0060, January 2012 - December 2013

Respirator Use Evaluation Worksheets, 10/9/13

DOP Test Log Sheets, 3/10/14, 3/20/14, 3/24/14

Breathing Air Analysis Results, Scott Revolve 5016 Compressor, 6/1/13, 8/26/13, 12/6/13,

2/18/14

Breathing Air Analysis Results, U2 Containment Breathing Air, 3/14/13

Breathing Air Analysis Results, U2 Service Air Compressor 1, 12/6/13

Breathing Air Analysis Results, U2 Service Air Compressor 2, 6/1/13

Breathing Air Analysis Results, Hypress FTB Compressor, 6/1/13, 8/26/13, 12/6/13, 2/18/14

Breathing Air Analysis Results, U1 Service Air Compressors 2 & 3, 2/18/14

Breathing Air Analysis Results, U1 Service Air, 8/26/13

Breathing Air Analysis Results, U1 Equipment Hatch Compressor, 3/21/14

Laboratory Report Compressed Air/Gas Quality Testing, Scott Revolve 5016 Compressor,

2/12/13

Laboratory Report Compressed Air/Gas Quality Testing, Hypress FTB Compressor, 2/12/13

List of Maintenance Personnel with SCBA Qualification Assigned, 2/28/14

List of Operations Personnel with SCBA Qualification Assigned, 2/28/14

CAP Documents Fleet-HP-2013, Nuclear Oversight Audit of Health Physics, 7/15/13

CR 617112

CR 647987

CR 695027

Attachment

Section 2RS4: Occupational Dose Assessment

Procedures, Guidance Documents, and Manuals

NMP-HP-107-001, "Instructions for Retrieving, Printing and Updating Individual Radiation

Exposure Records", Ver.1.0

NMP-HP-105, "Comparisons of OSLD and ED Dosimetry Results", Ver. 1.1

NMP-HP-106, "Investigating of Exposures Exceedi ng Fleet Administrative Limits", Ver. 1.0

NMP-HP-103, "Skin Dose Assessment", Ver. 1.1

NMP-HP-100, "Bioassay Program", Ver. 1.1

NMP-HP-101, "In-Vivo Bioassay and Internal Dose Assessment", Ver. 3.0

NMP-HP-102, "In-Vitro Bioassay," Ver. 1.1

NMP-HP-201, "Personnel Dosimetry Program," Ver. 1.1

NMP-HP-204, "Use and Calibration of Whole Body Counters," Ver. 1.3

Records and Data

NVLAP Certification of Accreditation to ISO/IEC 17025:2005, for Lab Code:100551-0, dated

2/13/2013.

Vogtle Alpha Plant Characterization Study 2011 Update

Canberra Report of Performance Testing Results for Nuclear Enterprises (NE) Model SPM

904B/906 Personnel Portal Monitor, May 18, 2012

Personnel Contamination Events/Personnel Contamination Reports (PCE/PCR) Logs, 2/2012 -

3/2014

EDE & NRC Form 5 Calculations for Steam Generator Multibadging Jobs entry made on

3/25/14; Multibadge RCA Authorization/Worksheets

NMP-HP-109 Data Sheets, Investigation of Lost, Damaged or Malfunctioning Personnel

Dosimetry, for occurrence on 3/18/2014

NMP-HP-109 Data Sheet 2, Investigation of Lost, Damaged or Malfunctioning Personnel

Dosimetry, for occurrence on 6/12/2013

CAP Documents

CR 541097

CR 585435

CR 610472

CR 746418

CR 748897

Section 2RS5: Radiation Monitoring Instrumentation

Procedures, Guidance Documents, and Manuals

43802-C, "Calibration of Gamma Standards", Ver. 12.4

NMP-HP-700, "Radiation Protection Instrumentation Program," Ver. 1.0

NMP-HP-701, "Daily Instrumentation Source Checks," Ver. 1.3

NMP-HP-719, "Operation and Calibration of the CANBERRA

ARGOS-5AB Exit Monitor",

Ver. 2.0

NMP-HP-718, "Operation and Calibration of the CANBERRA

GEM-5 Gamma Exit Monitor",

Ver. 1.0

NMP-HP-709, "Calibration of the Small Article Monitor (SAM-12)", Ver. 1.0

NMP-HP-708, "Operation and Calibration of the MGPI Telepole Instrument", Ver. 3.0

43693-C, "Operation and Use of the JL Shepard Model 89-400 Calibrator", Ver. 2.2

Attachment Records and Data Work Order SNC551063, RMSOOS 1-RE003 Out of Service

Work Order SNC405774, SGBD to MN Cond Rad Mon Ch CAL 1RE0021-18M, 2/28/13

Work Order SNC405890, Plant Vent Post Accident COT 1RE12444C-18M, 8/20/12

System Health Report, Unit 1 1609-R

ad Monitoring System, 7/1/2013-9/30/2013

Fleet-HP-2013, Nuclear Oversight Audit of Health Physics, July 15, 2013

43689-C Data Sheet 1, Calibration of the Small Article Monitor, Rev 7, for

SAM-11 VEGP

1. 1151, 5/23/2012 & 5/22/2014

NMP-HP-708 Data Sheet1, Telepole Gamma Calibration, SN#

VEG-HP-1511 3/04/14

43635-C Data Sheet 2, AMS High Voltage and Flow Calibration, SN#

VEGP-1450, 3-13-14 & 3-

6-13

NMP-HP-703 Data Sheet 1, Calibration Sheet,

RO-20 SN#

VEGP-HP-1017 03-4-14 & 03-5-13

43658-C Data Sheet 1, Air Sampler Calibration Sheet, SN#

VEGP-HP-1372 11-26-13 & 11-28-

NMP-HP-719, "Operation and Calibration of the CANBERRA

ARGOS-5AB Exit Monitor" Data

Sheet 1, ARGOS 5AB Calibration Certificate, 12-3-13

CAP Documents

TE 710894

CR 713514

CR 745425

CR 765241

CR 779661

CR 785178

Section 4OA1: Performance Indicator (PI) Verification

Procedures, Guidance Documents, and Manuals

00163-C, NRC Performance Indicator and Monthly Operating Report Preparation and Submittal, Ver. 14.6

Records and Data Liquid Effluent Release Permits L-20131221-239-B and L-20140227-035-B

Gaseous Effluent Release Permits G-20131231-002-B and G-20140222-045-B

CAP Documents

CR 617317

CR 654735

CR 700923

CR 723420

Section 4OA2: Identification and Resolution of Problems

Condition Reports

CR 648248; Calculation Error Affects EAL Setpoints for AS1 and AG1

CR 648345; Revise Emergency Plan and EPIP to correct EAL RS1 and RG1 error

CR 650353; Perform Apparent Cause Determination on Calculation Error

Attachment Documents: Southern Co. letter

NL-13-1979 to NRC, Emergency Plan Revision 60, dated September 24,

2013

Apparent Cause Determination Report, Calculation Errors Resulted in Incorrect EAL Setpoints, July 1, 2013

Documentation of Engineering Judgment

DOEJ-VXSNC648248-M001, Corrected Emergency

Action Level Set Points for RS1 and RG1 for Plant Vogtle, 5/31/2013

Procedures

91001-C, Emergency Classification and Implementing Instructions, Rev. 29

NMP-GM-002-001, Corrective Action Program Instructions, Ver. 31.1

NMP-GM-002-007, Apparent Cause Determination Instruction, Ver. 10

Section 4OA5: Other Activities

Condition Reports

CR 687752, 2B EDG Operability Assessment -

CAR 195200

NOTICE OF VIOLATION

Southern Nuclear Operating Company, Inc

Docket No. 50-424, 50-425

Vogtle Electric Generating Plant

License No.

NPF-68,

NFP-81

EA-14-112

During an NRC inspection completed on June 30, 2014, one violation of NRC requirements was

identified.

In accordance with the NRC Enforcement Policy, the violation is listed below:

CFR Part 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 requirements in Appendix E of this part and the standards in 10 CFR

50.47(b)

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 plans call for reliance on information

provided by facility licensees for determinations of minimum initial offsite response

measures.

CFR Part 50, Appendix E, Section

IV.B., Assessment Actions, requires the 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, and the Commission.

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 March 20, 2008, until May 31, 2013, the licensee failed to

maintain the effectiveness of their emergency plan.

Specifically, the licensee failed to

maintain a standard emergency classification scheme which included facility effluent

parameters in that effluent parameter classification threshold values for RG1 (General

Emergency) and RS1 (Site Area Emergency) were significantly non-conservative.

These

monitors were being relied upon to continuously assess the impact of the release of

radioactive materials as well as provide criteria for determining the need for notification and

participation of local and State agencies.

This violation is associated with a White SDP finding.

The NRC has concluded that information regarding:

1) the reason for the violation; 2) the

actions planned or already taken to correct the violation and prevent recurrence; and, 3) the

date when full compliance was achieved, is already adequately addressed on the docket in

Inspection Report No. 05000424/2014003 and 05000425/2014003.

However, you are required

to submit a written statement or explanation pursuant to 10 CFR 2.201 if the description therein

does not accurately reflect your corrective actions or your position.

In that case, or if you choose to respond, clearly mark your response as a "Reply to a Notice of Violation,

EA-14-112,"

and send it to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington,

DC 20555-0001 with a copy to the Regional Administrator, Region II, and a copy to

the NRC Resident Inspector, within 30 days of the date of the letter transmitting this Notice.

EA-14-112

If you choose to respond, your response will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html.

Therefore, to

the extent possible, the response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.

In accordance with 10

CFR 19.11, you may be required to post this Notice within two working

days of receipt.

Dated this 6

th day of August, 2014.