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:	Joel Munday 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 ust 6, 2014

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 stations emergency plan and associated implementing procedures, and performing extent of condition/cause reviews throughout the Southern Nuclear Operating Company fleet. Following the internal review process, the revised emergency plan and associated implementing procedure were provided to the NRC.

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

95001 Supplemental Inspection for One or Two White Inputs in a Strategic Performance Area, to review the licensees 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 NRCs Agency Rules of Practice and Procedure, a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs 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

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 NRCs Enforcement Policy dated January 28, 2013. The NRCs 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 licensees 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,

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 licensees 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 transformers) 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 stations 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 licensees 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 backup 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 determined 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 feedwater system and the train C turbine-driven auxiliary feedwater system during 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 feedwater system and the train C turbine-driven auxiliary feedwater system during 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 licensees 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 structures, systems, and components susceptible to flooding. The inspectors verified that plant design features and plant procedures for flood mitigation were consistent with design requirements and internal flooding analysis assumptions. The inspectors also assessed the condition of flood protection barriers and drain systems. In addition, the inspectors verified the licensee was identifying and properly addressing issues using the corrective action program. Documents reviewed are listed in the Attachment.

• Unit 1 residual heat removal (RHR) and containment spray (CS) pump rooms (both trains) in auxiliary building

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

.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 licensees 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 licensees 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 licensees identification, assessment, and characterization of the problems as well as their corrective actions for returning the equipment to a satisfactory condition. The inspectors also interviewed system engineers 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 maintenance 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 licensees 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 licensees risk assessments and implementation of risk management actions. The inspectors also verified that the licensee was identifying and resolving problems with assessing and managing maintenance-related risk using the corrective action program. Additionally, for maintenance resulting from unforeseen situations, the inspectors assessed the effectiveness of the licensees planning and control of emergent work activities.

Documents reviewed are listed in the Attachment.

• Unit 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 licensees evaluations. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. Additionally, the inspectors reviewed a sample of corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with operability evaluations.

Documents reviewed are listed in the Attachment.

• CR 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 licensees performance against criteria established in the licensees procedures. Additionally, the inspectors attended the post-exercise critique to assess the licensees 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 areas 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 monitor (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 licensees 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 licensees 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 workers 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 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 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 means 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 licensees 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 associated with ALARA program implementation.

The inspectors evaluated the licensees 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees neutron dosimetry program including 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 licensees 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 licensees 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, continuous air monitors (CAMs), PCMs, SAMs, PMs, and liquid and gaseous effluent monitors, a WBC, count room equipment, and portable survey instruments. The inspectors 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, ARGOSTM-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 setpoint 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 licensees 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 licensees 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 licensees computerized corrective action database.

.2 Semi-Annual Trend Review

a. Inspection Scope

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

b. Findings 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 licensees actions:

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

• evaluation and disposition of operability/reportability issues

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

• classification and prioritization of the problem

• identification of root and contributing causes of the problem

• identification of any additional condition reports

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

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 licensees 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 licensees 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 licensees ability to make timely and accurate General Emergency and Site Area Emergency classifications based on the abnormal radiological initiating condition, in that decision-makers would have to rely on other means to classify the event (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 licensees 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 licensees 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 minimum initial offsite response measures.

10 CFR Part 50, Appendix E, Section IV.B., Assessment Actions, requires that means to be used for determining the magnitude of, and for continuously assessing the impact

of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and state agencies, the Commission, and other federal agencies.

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

Contrary to the above, from 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 classifications RG1 (General Emergency) and RS1 (Site Area Emergency) contained effluent radiation monitor threshold values significantly greater than appropriate. These monitors were being relied upon to determine the magnitude and for continuously assessing the impact of the release of radioactive materials, as well as providing criteria for determining the need for notification and participation of local and state agencies. Following review by a Significance Enforcement Review Panel and NRC management, the violation was determined to meet IMC 0305, Section 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 licensees 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 licensees 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 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 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 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 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees 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 NRCs 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 licensees staff. The inspectors confirmed that proprietary information was not provided 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

05000424/2014003-01 NCV Inadequate Maintenance Procedure Results in a Failed MSIV and a Manual Reactor Trip (Section 1R12)

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

05000425/2014003-04 NCV Inadequate Maintenance Procedures and Usage Results in a Failed MFRV and an Automatic Reactor Trip (Section 4OA3.1)

Closed

05000425/2014-001-00 LER Automatic Reactor Trip Due to Low Steam Generator Level (Section 4OA3.1)

05000424/2014-002-00 LER Manual Reactor Trip due to Main Steam Isolation Valve Failure (Section 4OA3.2)

05000425/2013007-02 URI Failure to Identify and Correct Potential Emergency Diesel Generator 2B Inoperability Following Failed Surveillance Testing (Section 4OA5)