IR 05000369/2014005

IR 05000369/2014005, IR 05000370/2014005; 10/01/2014 - 12/31/2014; McGuire Nuclear Station, Units 1 and 2; Fire Protection and Refueling and Other Outage Activities

ML15034A580
Person / Time
Site:	McGuire, Mcguire
Issue date:	02/03/2015
From:	Frank Ehrhardt NRC/RGN-II/DRP/RPB1
To:	Capps S Duke Energy Carolinas
References
IR 2014005
Download: ML15034A580 (49)

Text

UNITED STATES ruary 3, 2015

SUBJECT:

MCGUIRE NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000369/2014005 AND 05000370/2014005

Dear Mr. Capps:

On December 31, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your McGuire Nuclear Station Units 1 and 2. On January 13, 2015, the NRC inspectors discussed the results of this inspection with you and other members of your staff.

Inspectors documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented two findings of very low safety significance (Green) in this report.

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, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the McGuire Nuclear Station. Also, 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 McGuire Nuclear Station.

In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRC's Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of 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/

Frank Ehrhardt, Chief Reactor Projects Branch 1 Division of Reactor Projects Docket Nos.: 50-369, 50-370 License Nos.: NPF-9, NPF-17

Enclosure:

NRC Integrated Inspection Report 05000369/2014005 and 05000370/2014005 w/Attachment - Supplemental Information

REGION II==

Docket Nos.: 50-369, 50-370 License Nos.: NPF-9, NPF-17 Report No.: 05000369/2014005, 05000370/2014005 Licensee: Duke Energy Carolinas, LLC Facility: McGuire Nuclear Station, Units 1 and 2 Location: Huntersville, NC 28078 Dates: October 1, 2014, through December 31, 2014 Inspectors: J. Zeiler, Senior Resident Inspector R. Cureton, Resident Inspector W. Loo, Senior Health Physicist (Sections 2RS2, 2RS4, and 4OA1)

M. Meeks, Senior Operations Engineer (Section 1R11)

A. Nielsen, Senior Health Physicist (Sections 2RS1, 2RS3, and 4OA1)

R. Patterson, Acting Project Engineer (Section 4OA5.2)

T. Fanelli, Reactor Inspector (Section 4OA5.2)

J. Rivera, Health Physicist (Sections 2RS5)

A. Vargas, Reactor Inspector (Sections 1R08, 4OA3 and 4OA5.3)

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

SUMMARY OF FINDINGS

IR05000369/2014005, IR05000370/2014005; 10/01/2014 - 12/31/2014; McGuire Nuclear

Station, Units 1 and 2; Fire Protection and Refueling and Other Outage Activities.

The report covered a three month period of inspection by the resident inspectors and seven regional inspectors. Two Green findings, which were determined to involve non-cited violations (NCVs) of NRC requirements, were identified. The significance of inspection findings are indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP), dated June 2, 2011.

Cross-cutting aspects are determined using IMC 0310, Aspects Within The Cross-Cutting Areas, dated December 19, 2013. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated July 9, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision, Revision (Rev.) 5.

Cornerstone: Mitigating Systems

• Green: An NRC-identified Green NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for the licensees failure to adequately control fire ignition sources in the Unit 1 and Unit 2 exterior doghouses in accordance with the FPP requirements of Nuclear System Directive (NSD)-

313, Control of Transient Fire Loads. Specifically, temporary electric portable heaters were energized for several days without implementing required hourly fire watches, locating the energized heaters greater than prescribed separation distances from safety-related equipment, and preventing other transient combustible materials from being located near the heaters. The licensee placed this issue into their corrective action program (CAP) and took corrective actions to de-energize the heaters, distance the heaters away from safety-related feedwater isolation valve electrical cables, and remove unnecessary transient combustibles from the area.

The failure to control fire ignition sources in accordance with NSD-313 was a performance deficiency (PD) . The PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and adversely affected the cornerstone objective in that, a fire could have affected nearby safety-related feedwater isolation valve electrical cables which provide a shutdown mitigation function. The finding was determined to be of very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown condition. This finding had a cross cutting aspect of teamwork in the human performance area because individuals failed to effectively communicate and coordinate their activities to ensure that the temporary heaters were energized following prescribed fire protection control measures and written instructions (H.4). (Section 1R05)

Cornerstones: Mitigating Systems and Barrier Integrity

• Green: An NRC-identified Green NCV of Technical Specification 5.4.1.a, Procedures, was identified for the failure to properly implement containment cleanliness and material control closeout procedures in accordance with procedure PT/1A/4600/003F, Containment

Cleanliness and ECCS Operability Inspection, prior to entering Mode 4, following the Unit 1 refueling outage. Specifically, a large amount of unanalyzed general loose debris, as well as scaffolding with aluminum walkboards and fibrous lead blankets, were left in containment that could either contribute to emergency core cooling system (ECCS) recirculation sump screen blockage or containment hydrogen generation during design basis accidents. The licensee placed this issue into their CAP and took corrective actions to remove the loose debris and unanalyzed materials and performed re-inspections of containment to identify any additional loose debris or unanalyzed materials left in containment.

The failure to perform an adequate containment cleanliness and material control closeout following the Unit 1 refueling outage in accordance with procedure PT/1/A/4600/003F was a PD. The PD was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective in that, loose debris in containment could result in the debris being transported to the ECCS recirculation sump screens in the event of design basis accident and adversely affect the sump performance. In addition, the PD was associated with the configuration control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective in that, the failure to control scaffolding that contained unanalyzed amounts of aluminum in containment challenged the existing analysis for containment aluminum inventory limitations. The finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of safety function of the ECCS sumps, was not safety significant due to external events, and no actual open pathway in the physical integrity of containment occurred. The finding had a cross-cutting aspect of field presence in the human performance area because the licensee failed to ensure that adequate supervisory and management oversight of the containment closeout process was conducted to ensure proper performance of procedure PT/1/A/4600/003F prior to entering Mode 4 (H.2). (Section 1R20)

REPORT DETAILS

Summary of Plant Status

Unit 1 began the inspection period shut down for a refueling outage. The unit was returned to 100 percent rated thermal power (RTP) on November 23, 2014, and operated at essentially full power until December 12, when a planned power reduction to 47 percent RTP was intiated to repair a main feedwater pump turbine bearing lube oil flow issue. The unit was returned to 100 percent RTP on December 15, and operated at essentially full power for the remainder of the inspection period.

Unit 2 operated at approximately 100 percent RTP for the entire inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

a. Inspection Scope

Readiness for Seasonal Extreme Weather Conditions: The inspectors reviewed the effectiveness of the licensees cold weather protection program pertaining to their preparations for seasonal cold weather conditions experienced during the inspection period. The inspectors discussed the licensees cold weather program with the assigned plant system engineer and verified that the licensee had implemented their cold weather preparation procedures. The inspectors walked down freeze protection equipment associated with the Unit 1 and Unit 2 refueling water storage tanks, interior/exterior doghouses (containing main steam/feedwater piping), and main feedwater flow transmitter compartments. These equipment/areas were selected because their importance to safety-related functions could be affected by adverse weather (freezing conditions). The inspectors observed plant conditions and evaluated those conditions against the criteria in the monthly equipment freeze protection checkout procedure.

Documents reviewed are listed in the Attachment.

Readiness for Impending Adverse Weather Conditions: The inspectors reviewed the effectiveness of the licensee's implementation of severe weather program response actions for a tornado watch issued on October 14, 2014, for Mecklenburg County. This included responding to the control room following announcement of the condition on the plant public address system and observing licensee actions required by emergency procedure RP/0/A/5700/006, Natural Disasters, Rev. 28. The inspectors verified the licensee implemented appropriate actions to protect personnel and mitigating system equipment from adverse weather effects in accordance with the procedure.

Readiness to Cope with External Flooding: The inspectors evaluated aspects of the licensees external flood protection features. This included a walkdown of the northern earthen dike extension of the Cowans Ford Dam and the condition of the site yard drains inside the protected area. The inspectors assessed the structural integrity and general condition of the earthen dike which is designed to protect safety-related facilities from flooding by Lake Norman. The inspectors verified that yard drains were intact and free of debris to prevent potential drain blockage. The inspectors reviewed the licensees CAP database to ensure that the licensee was identifying flood protection issues and resolving them commensurate with their significance. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

1R04 Equipment Alignment

a. Inspection Scope

Partial Walkdowns: The inspectors performed a partial walkdown of the following three systems to assess the operability of redundant or diverse trains and components when safety equipment was inoperable or degraded. The inspectors focused on discrepancies that could impact the function of the system and potentially increase risk. The inspectors reviewed applicable operating procedures and walked down control systems components to verify selected breakers, valves, and support equipment were in the correct position to support system operation. Documents reviewed are listed in the

.

• Unit 1 safety injection system flowpath alignment during refueling outage lowered inventory conditions

• 1A and 1B motor driven auxiliary feedwater (MDCA) pumps following the refueling outage while the turbine driven auxiliary feedwater (TDCA) pump was unavailable

• 1A emergency diesel generator (EDG) while the 1B EDG was out of service for planned preventive maintenance Complete System Walkdown: The inspectors conducted a detailed review of the Unit 1 residual heat removal (ND) system. To determine the correct system alignment, the inspectors reviewed operating procedures, drawings, and the updated final safety analysis report (UFSAR). Items reviewed during the inspection included:

(1) verification of correct valve positions and leak tightness of valve packing;

(2) availability of electrical power;

(3) correct labeling, cooling, and lubrication of system components;

(4) correct installation and functionality of hangers and supports;

(5) proper configuration and functionality of essential support systems;

(6) adequacy of area housekeeping and control of transient combustibles; and

(7) accuracy and appropriateness of component tagging and clearances. To determine the effect of outstanding design issues on the operability of the system, the inspectors reviewed the operator workaround list, the temporary modification list, system health reports, and other outstanding items tracked by the engineering department. In addition, the inspectors reviewed outstanding maintenance work requests/work orders and deficiencies that could affect the ability of the system to perform its function. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

1R05 Fire Protection

a. Inspection Scope

Fire Protection Walkdowns: The inspectors walked down accessible portions of the following five plant areas to determine if they were consistent with the UFSAR and the fire protection program for defense in depth features. The features assessed included the licensees control of transient combustible material and ignition sources, fire detection and suppression capabilities, firefighting equipment, and passive fire features such as fire barriers. The inspectors also reviewed the licensees compensatory measures for fire deficiencies to determine if they were commensurate with the significance of the deficiency. The inspectors reviewed the fire plans for the areas selected to determine if they were consistent with the fire protection program and presented adequate firefighting strategies. Documents reviewed are listed in the

.

• Unit 1 and Unit 2 auxiliary cable room (fire areas 19 and 20)

• Unit 1 upper and lower containment (fire area 32)

• Unit 1 reactor trip switchgear and motor generator room (fire area 22)

• Standby shutdown facility (SSF) (fire area YARD)

• Unit 1 and Unit 2 exterior doghouses (fire areas 30 and 31)

b. Findings

Introduction:

An NRC-identified Green NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for the failure to adequately control fire ignition sources in the Unit 1 and Unit 2 exterior doghouses. Specifically, temporary electric portable heaters were energized for several days without implementing required hourly fire watches, locating the energized heaters greater than prescribed separation distances from safety-related equipment, and preventing other transient combustible materials from being located near the heaters while energized.

Description:

On November 18, 2014, while verifying that appropriate FPP administrative controls were implemented for operating temporary 600 volt AC portable heaters in the Unit 1 and Unit 2 exterior doghouses for cold weather contingencies, the inspectors learned from interviews with the on-shift auxiliary operator assigned to the area, that the heaters had been energized since the previous morning. The inspectors questioned whether a fire watch had been established when the heaters were energized and learned that no fire watch had been in place during the period. The inspectors reviewed the fire impairment log and noted that Fire Impairment IMP-MC-2014-00421, which was generated relative to staging the heaters, required an hourly fire watch when the heaters were energized; however, this impairment had not been implemented. In addition to the fire impairment, the inspectors noted that a transient fire load evaluation form (NSD Form 313-1) was completed in accordance with Nuclear System Directive (NSD)-313, Control of Transient Fire Loads, and was posted to the wall in the area where each heater was staged. These evaluation forms were approved by the fire protection engineer on November 5 and allowed the staging of the portable heaters with several supplemental fire protection measures that were prescribed. These fire protection measures included the following:

• A qualified hourly fire watch patrol shall be implemented when the heaters are energized,

• Fire detection system shall be functional,

• Electrical extension cords shall be evaluated for proper amperage rating required for each heater,

• Separation distances between electrical equipment shall be maintained,

• Keep heater areas free of combustibles, and,

• Should the fire detection system become nonfunctional, a continuous fire watch shall be implemented while the heaters are energized.

Contrary to the above fire protection measures, in addition to the failure to implement an hourly fire watch, the inspectors identified that several of the heaters were in close proximity (less than three feet) to safety-related cables associated with the main feedwater isolation valves and each heater had flammable materials attached near the suction of the blower fan including a heater instruction manual contained in a plastic bag and a 20-30 foot section nylon rope, that in some cases, was laying on top of the heater housing. The licensee took immediate actions to de-energize the four heaters.

Subsequently, the combustible material attached to the heaters was removed and the heaters were moved further away from the main feedwater isolation valve cables. The licensees investigation into the how the heaters were energized on November 17, identified a communication weakness between a maintenance technician and his supervisor occurred resulting in the technician energizing all four heaters. The actions were implemented without any work order or other documentation in hand, which might have prompted the need to discuss the actions with operations and identify that a fire watch was required.

Analysis:

The licensees failure to adequately control fire ignition sources in the Unit 1 and Unit 2 exterior doghouses in accordance with procedure NSD-313, Control of Transient Fire Loads, was a PD. The PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, in that, uncontrolled fire ignition sources such as portable heaters could result in fires that could impact safety-related equipment in the area. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated June 2, 2011, Attachment 4, Initial Characterization of Findings, dated June 19, 2012. Using the guidance in IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, dated September 21, 2013, the finding was assigned a category of fire prevention and administrative controls. The inspectors determined the finding to be of very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown condition, in that, a postulated fire in the exterior doghouse did not present the possibility of impacting more than one train of safe shutdown equipment.

The finding had a cross cutting aspect of teamwork in the human performance area because individuals failed to effectively communicate and coordinate their activities to ensure that the temporary heaters were energized following prescribed fire protection control measures and written instructions (H.4).

Enforcement:

McGuire Nuclear Station Unit 1 and 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection, required the licensee to implement and maintain in effect all provisions of the approved FPP as described in Section 9.5.1 of the UFSAR. UFSAR Section 9.5.1.2.1, (General) Program Description, stated, in part, that FPP administrative controls are included in NSDs to manage control of flammable and combustible materials, hot work activities, and impairments to fire protection features.

The licensee implemented this requirement, in part, through procedure NSD-313, Control of Transient Fire Loads, which specifies conditions for completing a transient fire load evaluation when storing temporary equipment in a designated fire area via NSD Form 313.1, Transient Fire Load Approval Form. On November 5, 2014, a fire load evaluation was completed by a fire protection engineer via NSD Form 313.1 to allow temporary storage of portable heaters in the Unit 1 and Unit 2 exterior doghouses for additional freeze protection precautions. The evaluation required, in part, that when the heaters were energized, an hourly fire watch shall be maintained, separation distances between safety-related electrical equipment shall be maintained, and the heater area shall be free of combustibles. Contrary to the above, on November 18, 2014, the inspectors identified that the licensee did not adequately implement the FPP as required by NSD-313, in that, electrical heaters in the Unit 1 and Unit 2 exterior doghouses were energized without an hourly fire watch being established, adequate separation distances from safety-related feedwater isolation valve cables were not maintained, and combustible materials were left in close proximity to the heaters. The licensee took immediate actions to de-energize the four heaters. Subsequently, the combustible material attached to the heaters was removed and the heaters were moved further away from the main feedwater isolation valve cables. Because this finding was of very low safety significance and has been entered into the CAP as PIP M-14-11230, this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000369, 370/2014005-01, Failure to Adequately Control Transient Combustible Materials and Ignition Sources in Accordance with the Fire Protection Program)

1R06 Flood Protection Measures

a. Inspection Scope

Internal Flooding Reviews: The inspectors reviewed the UFSAR and the licensees flooding analysis to determine which plant areas were subject to internal flooding and contained safety-related equipment. The inspectors walked down the Unit 1 and Unit 2 turbine building, to determine whether the area configuration and flood protection barriers and equipment were consistent with the descriptions and assumptions described in UFSAR and licensee flooding analysis. The inspectors examined important flood protection equipment (i.e., flood barriers, sump pumps, and sump level instrumentation) to confirm that the equipment was being properly maintained in a state of functional readiness. The inspectors reviewed the operator actions contained in the licensees flood mitigation procedures to determine whether the desired results could be achieved within the times credited in the flooding analysis. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

1R07 Heat Sink Performance - Annual Resident Inspection

a. Inspection Scope

The inspectors selected the 1A component cooling water (KC) system and 1A EDG cooling water (KD) heat exchangers, based on their risk significance, and observed or reviewed aspects of the inspections, cleaning, and eddy current testing to determine whether the heat exchangers were ready and available to perform their intended functions as described in the UFSAR. The inspectors evaluated whether the frequency of inspections were sufficient to detect degradation prior to loss of heat removal capabilities below design requirements; that the inspection results were appropriately categorized against pre-established engineering acceptance criteria, including the impact of tubes plugged on the heat exchanger performance; that the licensee had developed adequate acceptance criteria for bio-fouling controls; and that the heat exchangers were properly reassembled with regard to end-bell orientation. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

a. Inspection Scope

Non-Destructive Examination Activities and Welding Activities: From September 22, through December 19, 2014, the inspectors conducted onsite and in-office reviews of the implementation of the licensees Inservice Inspection (ISI) Program for monitoring degradation of the reactor coolant system, emergency feedwater systems, risk-significant piping and components, and containment systems in Unit 1. The inspectors activities included a review of non-destructive examinations (NDE) to evaluate compliance with the applicable edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC),Section XI (Code of record: ASME Section XI 2007 Edition with the 2008 Addenda), and to verify that indications and defects (if present) were appropriately evaluated and dispositioned in accordance with the requirements of the ASME Code,Section XI, acceptance standards.

The inspectors directly observed and performed record review of the following NDEs mandated by the ASME Code to evaluate compliance with the ASME Code Section XI and Section V requirements, and, if any indications and defects were detected, to evaluate if they were dispositioned in accordance with the ASME Code or an NRC-approved alternative requirement.

Ultrasonic Testing (UT):

• Chemical and Volume Control Pipe to Valve, Weld No. 1NV1FW-53-27

• Intermediate Safety Injection 6 Stainless steel pipe to elbow, Weld No. 1NI-469-7

• Pressurizer Spray Nozzle to Head, Weld No. 1PZR-12

• Reactor Coolant Elbow to Pipe, Component No. Cold Leg Nozzle 1-3

• Reactor Coolant Elbow to Pipe, Component No. Cold Leg Nozzle 4-5

• Reactor Coolant Elbow to Pipe, Component No. Crossover Leg 1A, Nozzle No. 1-6

• Reactor Coolant Nozzle to Pipe, Component No. Cold Leg 1D Nozzle 4-1

• Reactor Coolant Pipe to Nozzle, Weld No. 1NC1F-1374

• Reactor Coolant Pipe to Nozzle, Weld No. 1NC1F-1377

• Residual Heat Removal 14 base metal exam, Component No. 1ND19A/1FW28

• Residual Heat Removal Pipe Component No. 1ND24-Tee

• Residual Heat Removal Pipe Tee Downstream of valve 1ND14, Component No.

1ND33-Tee

• Residual Heat Removal Pipe Tee Downstream of valve 1ND34 and Piping Elbow, Component No. 1ND34-Tee

• Residual Heat Removal Pipe Tee Downstream of valve 1ND4B, Component No.

1ND23-Tee

• Residual Heat Removal Pipe Tee toward elbow downstream of valve 1ND29, Component No. 1ND32-Tee

• Residual Heat Removal Pipe to Elbow, Weld No. 1ND1FW-32-9

• Residual Heat Removal Pipe to Flange, Weld No. 1ND1FW-32-7

• Residual Heat Removal Pipe to Tee, Weld No. 1ND1FW-32-5

• Residual heat Removal Tee, Weld No. 1ND-32

• Safety Injection 1B Cold Leg Pipe to Nozzle, Weld No. NC1F1493

• Safety Injection 1C Cold Leg Pipe to Nozzle, Weld No. NC1F1615

• Safety Injection 1D Cold Leg Nozzle No. 4-1

• Safety Injection Pipe to Elbow, Weld No. 1NI207A-4 Penetrant Testing (PT):

• Chemical and Volume Control System Rigid Support Welded attachment No.

1-MCA-NV-H15 Visual Testing (VT):

• Reactor Vessel Bottom Mounted Instrumentation The inspectors observed the welding activities referenced below and reviewed associated documents in order to evaluate compliance with procedures and the ASME Code. The inspectors reviewed the work order, repair and replacement plan, weld data sheets, welding procedures, procedure qualification records, welder performance qualification records, and NDE reports.

• Class 1, Refueling Water 1 inch split wedge gate valve replacement, Component No.

1-FW-87

• Class 1, Safety Injection Safety Injection 1B Cold Leg Pipe to Nozzle, Weld No.

NC1F1493

• Safety Injection 1C Cold Leg Pipe to Nozzle, Weld No. NC1F1615 During non-destructive surface and volumetric examinations performed since the previous refueling outage, the licensee did not identify any relevant indications that were analytically evaluated and accepted for continued service. Therefore, no NRC review was completed for this inspection procedure attribute.

Pressurized Water Reactor Vessel Upper Head Penetration Inspection Activities: For the Unit 1 vessel head, a bare metal visual (BMV) was examination required this outage pursuant to 10 CFR 50.55a. The inspectors reviewed portions of the Unit 1 BMV examination and reviewed NDE reports for vessel upper head penetration (VUHP) Nos.

15, 33, 58, 69 for the BMV examinations to determine if the activities, including the disposition of indications and defects, were conducted in accordance with the requirements of ASME Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). In particular, the inspectors evaluated if the required visual examination scope/coverage was achieved and limitations (if applicable) were recorded in accordance with the licensee procedures. Additionally, the inspectors evaluated if the licensees criteria for visual examination quality, and instructions for resolving interference and masking issues were consistent with 10 CFR 50.55a.

The licensee did not identify any relevant indications that were accepted for continued service during the BMV exam. Additionally, the licensee did not perform any welding repairs to the vessel head penetrations since the beginning of the last Unit 1 refueling outage. Therefore, no NRC review was completed for these inspection procedure attributes.

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

The inspectors also interviewed the BACCP owner, conducted an independent walkdown of containment to evaluate compliance with licensees BACCP requirements, and verified that degraded or non-conforming conditions, such as boric acid leaks, were properly identified and corrected in accordance with the licensees BACCP and CAP.

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

• PIP M-14-03582, Evidence of inactive leak, 04/09/14

• PIP M-14-04319, Steady stream of drops, 04/24/14

• PIP M-14-06197, Stationary film or drop, 06/24/14

• PIP M-14-07190, Boric Acid drop every 5 minutes, 08/11/14 The inspectors reviewed the following engineering evaluations completed for evidence of boric acid leakage to determine if degraded components were documented in the CAP.

The inspectors also evaluated corrective actions for any degraded components to determine if they met the ASME Section XI Code.

• MN20148010, 03/22/14

• MN20137637, 06/09/14

• MN20148195, 07/01/14

• MN20148211, 07/23/14 Steam Generator Tube Inspection Activities: The licensee did not perform steam generator tube inspection activities during the fall 2014 refueling outage. Therefore, the inspectors did not implement the inspection procedure attributes applicable to tube examination and repair activities. The inspectors reviewed the latest Condition Monitoring and Operational Assessment Report for Unit 1 to verify that the licensees evaluation of inspection results on the primary and secondary sides provided reasonable assurance that the tube integrity performance criteria would be met until the next scheduled inspection. The inspectors also verified that the planned tube inspection schedule was in accordance with the plants technical specifications.

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

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification (LOR) Program and Licensed Operator Performance

a. Inspection Scope

Quarterly Resident Inspector LOR Activity Review: On December 10, 2014, the inspectors observed operators in the plant simulator during a licensed operator requalification examination. The simulator examination scenario involved a failure of source range instrument 1NI-31 as well as a loss of grid and loss of residual heat removal while in shutdown conditions. The inspectors assessed overall crew performance, clarity and formality of communications, use of procedures, alarm response, control board manipulations, group dynamics and supervisory oversight. The inspectors observed the post-exercise critique to determine if the licensee identified deficiencies and discrepancies that occurred during the simulator training. Documents reviewed are listed in the Attachment.

Quarterly Resident Inspector Licensed Operator Performance Review: On November 14-15, 2014, the inspectors observed operators in the Unit 1 main control room during the performance of reactor startup and reactor physics testing following the refueling outage. The inspectors assessed the adequacy of overall crew performance, clarity and formality of communications, use of procedures, alarm response, control board manipulations, thoroughness of pre-job briefings, reactivity management controls, and supervisory oversight. Documents reviewed are listed in the Attachment.

Annual Review of Licensee Requalification Examination Results: On July 8, 2014, the licensee completed the comprehensive biennial requalification written examinations and the annual requalification operating examinations required to be administered to all licensed operators in accordance with 10 CFR 55.59(a)(2), Requalification Requirements, of the NRCs Operators Licenses. During the week of December 1, 2014, the inspectors performed an in-office review of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with NRC Inspection Procedure (IP) 71111.11, Licensed Operator Requalification Program. These results were compared to the thresholds established in Section 3.02, Requalification Examination Results, of IP 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the three issues listed below for items such as: 1) appropriate work practices; 2) identifying and addressing common cause failures; 3) scoping in accordance with 10 CFR 50.65(b) of the Maintenance Rule; 4) characterizing reliability issues for performance; 5) charging unavailability for performance; 6) balancing reliability and unavailability; 7) trending key parameters for condition monitoring; 8) classification and reclassification in accordance with 10 CFR 50.65(a)(1) or (a)(2); and 9) appropriateness of performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2) and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified as (a)(1). The inspectors performed a detailed review of the problem history and surrounding circumstances, evaluated the extent of condition reviews as required, and reviewed the generic implications of the equipment and/or work practice problem. Documents reviewed are listed in the

.

• PIP M-14-05421, Valve 2KC-57A failure to close during stroke time testing

• PIP M-14-08220, Containment isolation valve 1NV-7B failure to close during shutdown for refueling outage

• PIP M-14-11440, 1B motor generator set motor secured due to high vibrations

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensees risk assessments and risk management actions used to manage risk for the plant configurations associated with the four activities listed below. The inspectors assessed whether the licensee performed adequate risk assessments and implemented appropriate risk management actions when required by 10 CFR 50.65(a)(4). For emergent work, the inspectors verified that any increase in risk was promptly assessed, that appropriate risk management actions were promptly implemented, and that work activities did not place the plant in unacceptable configurations. Documents reviewed are listed in the Attachment.

• Yellow risk on Unit 1 for spent fuel pool cooling during full core offload and single train availability

• Yellow risk on Unit 1 for planned reactor coolant system (RCS) lowered inventory conditions following fuel reload, reactor vessel head installed, and A train safety-related equipment unavailable

• Orange risk due to emergent RCS draining to reduced inventory during Unit 1 refueling outage to repair pressure isolation check valve 1NI-60

• Yellow risk due to emergent replacement of the 1B motor generator set motor following failure due to high vibration

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors reviewed the five technical evaluations listed below to determine whether technical specification (TS) operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors reviewed any compensatory measures taken for degraded SSCs to determine whether the measures were in-place and adequately compensated for the degradation. For the degraded SSCs, or those credited as part of compensatory measures, the inspectors reviewed the UFSAR to determine whether the measures resulted in changes to the licensing basis functions, as described in the UFSAR, and whether a license amendment was required per 10 CFR 50.59. Documents reviewed are listed in the Attachment.

• PIP M-14-09401, Rejectable flaw identified during ultrasonic testing of Unit 1 residual heat removal pump suction piping

• PIP M-14-10857, During Mode 4 RCS pressure isolation valve testing, check valve 1NI-60 exhibited excessive leakage

• PIP M-14-10865, Operability evaluation of loose debris, scaffolding, and lead shielding left in containment after Mode 4 entry

• PIP M-14-11129, During 1A EDG load sequencer testing, the engine started unexpectedly

• PIP M-14-11230, Operability evaluation of energized portable electrical heaters in the Unit 1 and Unit 2 exterior doghouse without a fire watch

b. Findings

No findings were identified.

1R18 Plant Modifications

a. Inspection Scope

The inspectors reviewed the following permanent plant modification to verify the adequacy of the modification package and 10 CFR 50.59 screening. The modification was evaluated against the TS, UFSAR, and licensee design basis documents for the systems affected to ensure the modification did not adversely affect the availability, reliability, and functional capability of important SSCs. Documents reviewed are listed in the Attachment.

• Engineering Change (EC) 109071, Fukushima modification to add assured suction water source to auxiliary feedwater system

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the six post-maintenance tests listed below to determine if procedures and test activities ensured system operability and functional capability. The inspectors reviewed the licensees test procedures to determine if the procedures adequately tested the safety functions that may have been affected by the maintenance activities, that the acceptance criteria in the procedures were consistent with information in the applicable licensing basis and/or design basis documents, and that the procedures had been properly reviewed and approved. The inspectors also witnessed the tests and/or reviewed the test data to determine if test results adequately demonstrated restoration of the affected safety functions. Documents reviewed are listed in the

.

• Auxiliary feedwater suction supply isolation valve 1CA-162B functional testing following modification to add assured air supply controls

• RCS pressure isolation check valve 1NI-60 functional testing following emergent repair for excessive leakage

• Initial criticality and zero power physics testing following Unit 1 refueling outage

• 1A load sequencer testing following emergent repair of failed relay

• Unit 1 measurement uncertainty recapture power uprate testing following Unit 1 refueling outage

• 1B EDG functional testing following planned maintenance

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors conducted the following inspection activities associated with the Unit 1 refueling outage that completed November 17, 2014. Documents reviewed are listed in the Attachment.

• Observed activities to verify that the licensee maintained defense-in-depth commensurate with the outage risk control plan for key safety functions and applicable TS when taking equipment out of service

• Reviewed the licensees responses to emergent work and unexpected conditions to verify that resulting configuration changes were controlled in accordance with the outage risk control plan

• Periodically reviewed the setting and maintenance of containment integrity to establish that the reactor coolant system and containment boundaries were in place and had integrity when necessary

• Observed fuel handling operations during reactor core reload including review of the videotape core loading verification and alignment to verify that those operations and activities were being performed in accordance with TS and procedural guidance

• Observed aspects of the reactor vessel head reinstallation and upper internals to ensure the lifts were conducted in accordance the station procedures and heavy lift guidance

• Reviewed system lineups and/or control board indications to substantiate that TS, license conditions, and other requirements, commitments, and administrative procedure prerequisites for mode changes were met prior to changing modes or plant configurations

• Conducted containment walkdowns to inspect for overall cleanliness and material condition of plant equipment after the licensee completed their closeout inspection prior to restart

• Observed the approach to criticality, portions of reactor physics testing and power ascension activities

• Reviewed the items that had been entered into the CAP to verify that the licensee had identified outage related problems at an appropriate threshold

b. Findings

Introduction:

An NRC-identified Green NCV of TS 5.4.1.a, Procedures, was identified for the failure to properly implement containment cleanliness and material control closeout procedures prior to entering Mode 4, following the refueling outage. The improper closeout resulted in a large amount of unanalyzed general loose debris, as well as scaffolding with aluminum walkboards and fibrous lead blankets, left in containment that could either contribute to emergency core cooling system (ECCS) recirculation sump screen blockage or containment hydrogen generation during design basis accidents.

Description:

On November 5, 2014, the licensee completed surveillance procedure PT/1/A/4600/003F, Containment Cleanliness and ECCS Operability Inspection. The main purpose of this procedure was to ensure that no loose debris/materials was left in containment following the refueling outage that could be transported to the containment ECCS recirculation sump and cause restriction of the ECCS pumps suction during design basis accidents. In addition, the procedure requires that any temporary materials not removed from containment (such as scaffolding, lead blankets, etc.) be identified and evaluated to be acceptable to be left in containment with unit operation in Modes 1-4.

Items evaluated and allowed to be left in containment were required to be logged in PT/1/A/4600/003F, Enclosure 13.1, Loose Material and Equipment Data Sheet.

On November 7, following the licensees entry into Mode 4, the inspectors reviewed the completed PT/1/A/4600/003F, including the list of materials approved to be left inside containment via Enclosure 13.1 and performed a walkdown of portions of the Unit 1 lower containment to verify the adequacy of the licensees containment closeout controls. During initial inspection inside the crane wall of the B and C RCS loops, the inspectors found a large amount of outage related general loose debris items that had not been removed as required by procedure PT/1/A/4600/003F. Items included cloth towels, gloves, duct tape, tie-wraps, paper, tools with nylon lanyards, flashlights, foam knee pads, and various other forms of small plastic. In addition, the inspectors identified scaffolding below the B reactor coolant pump that was not recalled to be on Enclosure 13.1 of PT/1/A/4600/003F. Prior to conducting further containment inspections, the inspectors exited containment to confirm whether the scaffolding identified was actually on the approval sheet and to discuss with licensee management the unusually large amount of loose debris that was found inside the crane wall.

The licensee subsequently initiated PIP M-14-10865 to address the discrepancies identified by the inspectors, confirmed that the scaffolding was not evaluated to be left in containment and took action to remove it, and decided to re-perform the containment cleanliness and material control inspections due to the large amount of loose debris found by the inspectors. During the licensees subsequent inspections, additional loose debris and unanalyzed material control items were identified and removed from containment. Particularly noteworthy of these discrepancies included six, 1-foot by 3-foot by1-inch thick, fibrous filled lead blankets with vinyl covering material that was used for temporary radiological shielding during the outage on a nuclear sampling piping line.

Following the licensees re-inspections, the inspectors completed the walkdown of all accessible areas of containment and verified the adequacy of the licensees corrective actions.

The licensee performed an engineering analysis of the loose debris left in containment, including the fibrous lead blankets, and determined that the cumulative volume and type of materials would not have caused the ECCS recirculation sump to be rendered inoperable had a design basis accident occurred. The licensee re-evaluated the piping stress analysis for the nuclear sampling system to take into consideration the additional loading from the six feet of lead blankets that were left installed on the piping and it was determined that the piping supports were capable of performing their design function with the additional load. The licensee determined that the scaffolding left in containment by mistake contained three, four foot length, walkboards made of aluminum that can generate hydrogen during design basis accident conditions. The licensee initially determined that the current containment aluminum (submergence) inventory limits would be exceeded with the additional aluminum from the walk boards. However, the calculation contained several areas that were conservative in the accounting of the aluminum inventory, e.g., valves containing aluminum that were included in the submergence category when the valves were known to be located at elevations that would not result in submergence. As a result of these further inventory refinements and corrections, the licensee determined that the actual aluminum submergence inventory limits would not have been exceeded with the addition of the scaffolding aluminum. The inspectors reviewed the licensees evaluations and agreed with the licensees conclusions regarding the impact of the discrepancies on the ECCS recirculation sump and containment post-accident pressure design.

Analysis:

The licensees failure to perform an adequate containment cleanliness and material control closeout following the Unit 1 refueling outage in accordance with procedure PT/1/A/4600/003F was a PD. The PD was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, loose debris in containment could result in the debris being transported to the ECCS recirculation sump screens in the event of design basis accident and adversely affect the sump performance. Also, the PD was associated with the configuration control attribute of the barrier integrity cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that the containment barrier design would protect the public from radionuclide releases during design basis accidents. Specifically, the failure to control scaffolding that contained unanalyzed amounts of aluminum in containment challenged the existing analysis for containment aluminum inventory limitations. Using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, for the mitigating systems and barrier integrity cornerstone areas, the inspectors determined the finding was of very low safety significance (Green)because it did not result in an actual loss of safety function of the ECCS sumps, was not safety significant due to external events, and no actual open pathway in the physical integrity of containment occurred. The finding had a cross-cutting aspect of field presence in the human performance area because the licensee failed to ensure that adequate supervisory and management oversight of the containment closeout process was conducted to ensure proper performance of procedure PT/1/A/4600/003F prior to entering Mode 4 (H.2).

Enforcement:

Technical Specification 5.4.1.a, Procedures, requires, in part, that written procedures be established, implemented, and maintained covering activities described in Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, Appendix A, February 1978. Section 2.a of Regulatory Guide 1.33, Appendix A, requires general operating procedures for plant startup from cold shutdown (i.e., Mode 5) to hot standby (i.e., Mode 4). Operating procedure OP/1/A/6100/SU-9, Mode 4 Checklist, Enclosure 4.1, Operations Checklist, Step 3.59, requires that a containment closeout inspection be performed in accordance with surveillance procedure PT/1/A/4600/003F, Containment Cleanliness and ECCS Operability, prior to Mode 4 entry. PT/1/A/4600/003F requires that no loose material be present in containment which could be transported to the containment ECCS recirculation sump without evaluation and all other materials left in containment shall be identified, evaluated for acceptability, and documented in Enclosure 13.1, Loose Material or Equipment Data Sheet. Contrary to the above requirements, on November 7, 2014, it was identified that a large amount of loose debris was left in containment without an evaluation that could be transported to the ECCS recirculation sump, and scaffolding with aluminum walkboards and fibrous lead blankets had not been removed or evaluated acceptable to remain in containment, and was not documented in 13.1 of PT/1/A/4600/003F, prior to Unit 1 entering Mode 4. The licensee took immediate corrective actions to remove the loose debris and materials identified and performed re-inspections of containment to identify any additional loose debris or unanalyzed materials left in containment. Because this finding is of very low safety significance and has been entered into the CAP as PIP M-14-10865, this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000369/2014005-02, Failure to Adequately Implement Containment Closeout Resulting in Loose Debris and Unanalyzed Materials Left in Containment)

1R22 Surveillance Testing

a. Inspection Scope

For the five surveillance tests identified below, the inspectors witnessed testing and reviewed the test data to determine if the SSCs involved in these tests satisfied the requirements described in the TS, the UFSAR, and applicable licensee procedures. In addition, the inspectors verified that the tests demonstrated that the SSCs were capable of performing their intended safety functions.

Surveillance Tests

• PT/1/A/4200/009A, Engineered Safety Features Actuation Periodic Test Train A, Rev. 142

• PT/1/A/4200/002C, Containment Closure, Rev. 83 In-Service Tests

• PT/1/A/4252/007, CA System Turbine Driven Performance Test, Rev. 85 Containment Isolation Valve Testing

• PT/1/A/4200/001C, Isolation Valve Leak Rate Test, Rev. 123 (Enclosure 13.40 for Penetration M-374)

Ice Condenser Systems Testing

• PT/0/A/4200/032, Periodic Inspection of Ice Condenser Lower Inlet Doors, Rev. 21

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Occupational Radiation Safety (OS) and Public Radiation Safety (PS)

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, high radiation areas (HRA), and very HRA (VHRA) established within the radiologically controlled area (RCA) of the auxiliary building, Unit 1 (U1) reactor containment building, independent spent fuel storage installation (ISFSI), and radioactive waste (radwaste)processing and 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, discrete radioactive particles, airborne radioactivity, neutron exposure, 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 U1 end-of-cycle 23 (1EOC23) refueling outage tasks, 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 locked HRA (LHRA) and VHRA locations and discussed changes to procedural guidance for LHRA and VHRA controls with health physics (HP)supervisors. The inspectors reviewed implementation of controls for the storage of irradiated material within the spent fuel pool. Established radiological controls were evaluated for selected 1EOC23 tasks including removal of reactor head ventilation duct work and cutting/welding activities on safety injection piping. In addition, the inspectors reviewed licensee controls for areas where dose rates could change significantly as a result of plant shutdown and refueling operations.

Through direct observations and interviews with licensee staff, inspectors evaluated occupational workers adherence to selected RWPs and HP technician (HPT) proficiency in providing job coverage. Electronic dosimeter (ED) alarm set points and worker stay times were evaluated against area radiation survey results for selected 1EOC23 job tasks. As part of IP 71124.04, inspectors reviewed the use of personnel dosimetry (extremity dosimetry, multibadging in high dose rate gradients, etc.). The inspectors also evaluated worker responses to dose and dose rate alarms during selected work activities.

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 the last two 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 evaluated the licensees stewardship of nationally tracked radioactive sealed sources and reviewed leak testing records.

Problem Identification and Resolution: The inspectors reviewed CAP documents associated with radiological hazard assessment and control. 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.

Radiation protection (RP) activities were evaluated against the requirements of UFSAR Section 12; TS Section 5; 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 and records reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2RS2 Occupational ALARA Planning and Controls

a. Inspection Scope

Work Planning and Exposure Tracking: The inspectors reviewed work activities and their collective exposure estimates for the previous Unit 2 (U2) end-of-cycle 22 (2EOC22) and current 1EOC23 outages. ALARA planning packages were reviewed for the following high collective exposure tasks to include removal of reactor head ventilation duct work and cutting/welding activities on safety injection piping. For the selected tasks, the inspectors reviewed established dose goals and discussed assumptions regarding the bases for the current estimates with responsible ALARA planners. 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 (trigger points) for additional ALARA review. Where applicable, changes to established estimates were discussed with ALARA planners and evaluated 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 2011 - 2013 and reviewed historical outage collective exposure trends on each unit. The inspectors reviewed historical dose rate trends during post shutdown crudburst/cleanup. Source term reduction initiatives were reviewed and discussed with cognizant licensee staff.

Radiation Worker Performance: Radiation worker performance was observed and evaluated as part of IP 71124.01 and is documented in Section 2RS1. While observing job tasks, the inspectors evaluated the use of remote technologies to reduce dose including teledosimetry and remote visual monitoring. This included activities associated with a radioactive filter move, upgrading of incore sump room from a HRA to a VHRA, and a U2 at power entry into lower containment.

Problem Identification and Resolution: The inspectors reviewed and discussed selected corrective action documents associated with ALARA program implementation. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with licensee procedures. 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, RP; TS Section 5.4, Procedures; 10 CFR Part 20; and approved licensee procedures. Documents and records reviewed are listed in the 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 engineering controls to mitigate airborne radioactivity during the 1EOC23 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 certificates. 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 cutting and welding activities on safety injection piping. Selected self-contained breathing apparatus (SCBA) units and negative pressure respirators (NPR) 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 discussed the use of powered air-purifying hoods and full-face NPR during 1EOC23 activities inside containment with HP staff. The inspectors also discussed training for various types of respiratory protection devices and interviewed radworkers and control room operators on use of the devices including SCBA bottle change-out and use of corrective lens inserts. The inspectors reviewed respirator qualification records for several main control room operators and emergency responder personnel in the maintenance and HP departments. In addition, inspectors evaluated qualifications for individuals responsible for testing and repairing SCBA vital components.

Problem Identification and Resolution: The inspectors reviewed CAP documents 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 UFSAR Section 12, TS Section 5; 10 CFR Part 20; the guidance in Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection; and applicable licensee procedures. Documents and records reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2RS4 Occupational Dose Assessment

a. Inspection Scope

External Dosimetry: The inspectors reviewed National Voluntary Laboratory Accreditation Program certification data (including thermoluminescent dosimeter testing for neutron, gamma, and beta exposures) and discussed program guidance for storage, processing, and results for active and passive personnel dosimeters currently in use.

Licensee procedures for shallow and deep dose assessments for workers with identified skin contaminations were reviewed and discussed. Comparisons of ED and personnel dosimeter data were reviewed and discussed in detail. In addition, inspectors reviewed procedural requirements for extremity dosimetry, multi-badging, and re-positioning of whole body dosimetry.

Internal Dosimetry: Program guidance (including derived air concentration-hour tracking), instrument detection capabilities, and assessment results for internally deposited radionuclides were reviewed in detail. The inspectors discussed with licensee staff the stations passive monitoring program in use for routine in vivo (whole body counting) analyses. In addition, capabilities for collection and analysis of special bioassay samples were reviewed and evaluated.

Special Dosimetric Situations: The inspectors evaluated the licensees use of multi-badging, extremity dosimetry, and dosimeter relocation within non-uniform dose rate fields and discussed worker monitoring in neutron areas with licensee staff. This included activities associated with the 1EOC23 outage, U2 power entries, and ISFSI activities. The inspectors reviewed monitoring records for declared pregnant workers (DPW) since January 2013, and discussed DPW monitoring guidance with licensee staff.

In addition, the adequacy of shallow dose assessments for selected personnel contamination events were reviewed and discussed.

Problem Identification and Resolution: The inspectors reviewed and discussed selected CAP documents associated with occupational dose assessment. The inspectors evaluated the licensees ability to identify and resolve the identified issues in accordance with licensee procedures. The inspectors also discussed the scope of the licensees internal audit program and reviewed recent assessment results.

Occupational dose assessment program activities were evaluated against the requirements of FSAR Section 12, RP; TS Section 5.4, Procedures; 10 CFR Parts 19 and 20; RG 8.40, Methods for Measuring Effective Dose Equivalent from External Exposure; and approved licensee procedures. Documents and records reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation

a. Inspection Scope

Radiation Monitoring Instrumentation: During walk-downs of the liquid and gaseous radwaste systems in the auxiliary building, and the RCA exit point, the inspectors observed installed radiation detection equipment. These included area radiation monitors (ARMs), liquid and gaseous effluent monitors, PCMs, SAMs, and PMs. The inspectors observed the physical location of the components, noted the material condition, and compared sensitivity ranges with UFSAR details. Setpoint methodologies for selected radiation monitors were evaluated for correct alarm setpoint determination based on offsite dose calculation manual (ODCM) requirements.

In addition to equipment walk-downs, the inspectors observed functional checks alarm set-point testing of various portable and fixed detection instruments, including ion chambers, telepoles, PCMs, SAMs, PMs, and a whole body counter (WBC). For the portable instruments, the inspectors observed the use of a high-range calibrator and discussed periodic output value testing with a HPT. The inspectors reviewed calibration records and evaluated alarm setpoint values for selected PCMs, PMs, effluent monitors, ARMs, SAMs, and a WBC, including instruments used for post-accident monitoring. The radioactive source used to calibrate an effluent monitor was evaluated for traceability to national standards. Calibration stickers on portable survey instruments were noted during inspection of the storage area for ready-to-use equipment. The most recent 10 CFR Part 61 analysis for DAW was reviewed to determine if calibration and check sources were representative of the plant source term. The inspectors also reviewed countroom calibration records for a gamma spectroscopy germanium detector, a liquid scintillation detector, and alpha/beta smear counters.

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 and records reviewed are listed in the

.

Problem Identification and Resolution: The inspectors reviewed selected CAP documents in the area of radiological instrumentation. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with licensee procedures. Documents and records reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

The inspectors sampled licensee data to confirm the accuracy of reported PI data for the following six indicators. To determine the accuracy of the PI data reported during that period, the inspectors compared the licensees basis in reporting each data element to the PI definitions and guidance contained in NEI 99-02, Regulatory Assessment Indicator Guideline, Rev. 7, as well as licensee procedural guidance for reporting PI information. Documents reviewed are listed in the Attachment.

Barrier Integrity Cornerstone

• RCS Specific Activity (Units 1 and 2)

• RCS Leak Rate (Units 1 and 2)

The inspectors reviewed the PI results for October 1, 2013, through September 30, 2014. The inspectors compared the licensee reported PI data with records developed by the licensee that contained daily calculated values for RCS activity and leak rates.

The inspectors reviewed surveillance test records of measured RCS leak rates and control room logs documenting leak rate measurements and compared these to values submitted to the NRC. The inspectors interviewed the licensee personnel who were responsible for collecting and evaluating the PI data.

Occupational Radiation Safety Cornerstone

• Occupational Exposure Control Effectiveness The inspectors reviewed the occupational exposure control effectiveness PI results for the occupational radiation safety cornerstone from July 2013 through June 2014. The inspectors reviewed electronic dosimeter alarm logs and procedural guidance for identifying and reporting PI occurrences. Documents and records reviewed are listed in the Attachment.

Public Radiation Safety Cornerstone

• Radiological Control Effluent Release Occurrences The inspectors reviewed the radiological effluent technical specification/ODCM radiological effluent occurrences PI results from April 2013 through August 2014. The inspectors reviewed CAP documents, effluent dose data, and licensee procedural guidance for classifying and reporting PI events. Documents and records reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

a. Inspection Scope

Review of Items Entered into the Corrective Action Program: As required by IP 71152, Problem Identification and Resolution, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed screening of items entered into the licensees corrective action program. This was accomplished by reviewing copies of condition reports, attending some daily screening meetings, and accessing the licensees computerized CAP database.

Semi-Annual Review to Identify Trends: As required by IP 71152, Problem Identification and Resolution, the inspectors performed a review of the licensees CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review was focused on repetitive equipment issues, but also considered the results of daily inspector CAP item screenings, licensee trending efforts, and licensee human performance results. This review nominally considered the six month period of July 2014 through December 2014 although some examples expanded beyond those dates when the scope of the trend warranted. The review also included issues documented outside the normal CAP in major equipment problem lists, focus area reports, system health reports, self-assessment reports, and department PIP trending reports. The inspectors compared and contrasted their results with the results contained in the licensees latest quarterly trend reports. Documents reviewed are listed in the Attachment.

Annual Sample Reviews: The inspectors reviewed the issue listed below in detail to evaluate the effectiveness of the licensees corrective actions for important safety issues.

• PIP M-14-04872, 2B EDG voltage adjust would not reduce voltage while performing surveillance testing The inspectors assessed whether the issue was properly identified; documented accurately and completely; properly classified and prioritized; adequately considered extent of condition, generic implications, common cause, and previous occurrences; adequately identified root causes/apparent causes; and identified appropriate and timely corrective actions. The inspectors evaluated the licensee documents against the requirements of the licensees CAP and implementing procedures, and 10 CFR 50, Appendix B. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA3 Follow-Up of Events and Notices of Enforcement Discretion (NOED)

(Closed) Licensee Event Report (LER) 05000369/2014-002-00, Degraded Conditions Due To Rejectable Flaws in 1B and 1C Safety Injection Lines

a. Inspection Scope

The inspectors reviewed the LER described above, the associated root cause report (PIP M-14-9052), and discussed the issue with licensee staff. The licensee determined the direct cause of the 1B and 1C safety injection line rejectable flaws to be original construction deficiencies, which produced areas in the lines that contained residual stresses, and consequently made the areas susceptible to flaw initiation. The licensee also determined that a high cycle thermal fatigue condition was created by leakage from valve 1NI-3, which also contributed to the flaw initiation. The 1B and 1C rejectable flaws have been cut out, repaired, and replaced with new piping. All welds were examined satisfactorily before being returned to service. LER 05000369/2014-002-00 is closed.

b. Findings

No findings were identified.

4OA5 Other Activities

.1 Operation of an Independent Spent Fuel Storage Installation

a. Inspection Scope

Using the inspection guidance in IMC 60855.1, Operation of an ISFSI at Operating Plants, the inspectors reviewed revisions to the NAC-MAGNASTOR spent fuel storage cask FSAR and 10 CFR 72.212 Safety Evaluation Report (SER) made since the previous year to assess the impact of the changes on the licensees ISFSI program.

The inspectors observed the general condition of the ISFSI storage casks, the associated cooling vents, and the cask temperature monitoring system to verify that the casks were free of any structural damage, the vents were free of obstructions, and the temperature monitoring equipment was operating correctly. The inspectors reviewed fuel assembly physical inventory records to ensure the licensee was maintaining appropriate material control and accounting of spent nuclear fuel. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

.2 (Closed) Unresolved Item (URI) 05000369, 370/2013008-01, Fuel Manipulator Crane

Digital Modification

a. Inspection Scope

During the 2013 McGuire 50.59 modifications inspection (ML13353A115), the team identified a URI associated with a potential failure to conduct an adequate 10 CFR 50.59 evaluation for a digital modification to the fuel manipulator crane control system.

Specifically, engineering changes EC 77048, Unit 1 Manipulator Crane Upgrade Project, Rev. 18 and EC 77051, Unit 2 Manipulator Crane Upgrade Project, Rev. 15, were implemented to address reliability and obsolescence issues with the fuel manipulator cranes in the reactor and spent fuel building.

The existing analog controls were replaced with a digital programmable logic controller with a graphic user interface. The power and control systems were upgraded, including the position sensors, motor drives, control consoles, and wiring. The licensee performed a 10 CFR 50.59 evaluation in accordance with procedure NSD-209, 10 CFR 50.59 Process, Rev. 14, and determined the change could be implemented without prior NRC review and approval. Procedure NSD-209 requires the license to prescribe controls and methods for implementing the requirements of 10 CFR 50.59 consistent with NEI 96-07, Rev. 1, Guidelines For 10 CFR 50.59 Evaluations.

The issue was unresolved after reviewing the 10 CFR 50.59 evaluation, because the inspectors could not verify the licensees conclusions regarding the reliability and dependability of the software used to operate the manipulator crane. Specifically, the licensee did not address factors that affect digital system dependability in enough detail for the team to evaluate the conclusions.

NEI 96-07, Rev. 1, states, in part, that, In performing a 10 CFR 50.59 evaluation of a proposed activity, the evaluator must address the eight criteria in 10 CFR 50.59(c)

(2) to determine if prior NRC approval is required. Although the conclusion in each criterion may be simply yes, no or not applicable, there must be an accompanying explanation providing adequate basis for the conclusion. Consistent with the intent of 10 CFR 50.59, these explanations should be complete in the sense that another knowledgeable reviewer could draw the same conclusion.

b. Findings

The licensees failure to address factors that affect digital system dependability in enough detail for the team to evaluate the sites conclusions was a performance deficiency. This performance deficiency was considered to be not more than minor because the licensee was able to provide the team with reasonable assurance that any malfunction of the software would not lead to a more significant safety concern.

Specifically, the licensee provided the team a revised 50.59 evaluation highlighting the quality standards used to develop the software, and addressed the potential failure modes and effects of the software when operating the manipulator crane. In addition, the licensee provided the team with calculation MCC-1201.10-00-0010, Fuel Manipulator Crane Extended Analyses per PIP M13-10230 Software Hazard Coping Analysis, which addresses common cause failures, malfunctions, and operator actions to maintain the manipulator crane in a safe mode by pushing the emergency stop button which in turn disconnects all power to the crane preventing any movement.

Based on the teams review of the revised the 50.59 evaluation and the additional information provided by the licensee, the team concluded that prior NRC approval would not have been required. This URI is considered closed. The licensee entered this item into their corrective action program as problem investigation program (PIP) report M13-11029 to track the actions taken to address all of the teams additional observations.

.3 (Closed) URI 05000369/2014004-01, 1B/1C Reactor Coolant System Loop Safety

Injection Piping Flaws

a. Inspection Scope

In Inspection Report 05000369/2014004, the inspectors identified an unresolved item (URI) associated with flaws identified in the 1B/1C reactor coolant system loop safety injection piping welds. The licensee identified flaws with UT in the 1B and 1C cold leg safety injection pipe welds as part of their extent of condition from Unit 2 for Materials Reliability Program (MRP)-146, Thermal Fatigue. Further evaluation determined these flaws were a circumferential flaw with an axial component on the nozzle side for 1B, and an axial flaw from the centerline of the weld into the base metal for 1C. The licensee completed examinations on all welds included in the MRP-146 program and found them to be within the acceptance criteria. The licensee also removed and repaired the 1B and 1C nozzles. Welding of the new components has been examined and has passed all quality assurance examinations. The licensee determined that the flaws were a result of high cycle thermal fatigue and residual stresses created by construction deficiencies.

The inspectors reviewed documentation associated with the identification of the flaw, categorization, repairs and the metallurgical analysis performed. The inspectors determined that although flaws existed in the 1B/1C reactor coolant system loop safety injection piping welds, the licensee performed all required examinations, repairs and metallurgical analysis of the flaws. The licensee also implemented corrective actions to their inspection schedule and inspection techniques used for these type welds to better monitor thermal fatigue in these type welds and configuration. URI 05000369/2014004-01 is closed.

b. Findings

No findings were identified.

.4 (Closed) Temporary Instruction (TI) 2515/190, Inspection of the Licensees Proposed

Interim Actions as a Result of the Near-Term Task Force Recommendation 2.1 Flooding Reevaluation The inspectors verified the licensees interim actions will perform their intended function for flooding mitigation. The inspectors independently verified that the licensees proposed interin actions would perform their intended function for flooding mitigation.

Specific inspection activities included the following:

• Visual inspection of the flood protection features was performed if the flood protection feature was relevant,

• External visual inspection for indications of degradation that would prevent its credited function from being performed,

• Reasonable simulation, if applicable to the site, and,

• Flood protection feature functionality was determined using either visual observation or by review of other documents.

The inspectors verified that issues identified were entered into the licensees CAP.

4OA6 Meetings, Including Exits

On January 13, 2015, the resident inspectors presented the inspection results to Mr. Steven Capps and other members of his staff. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

B. Aderhold- ISI

B. Anderson, Superintendent of Operations

D. Black, Security Manager

D. Brenton, Maintenance Superintendent

S. Capps, Vice President, McGuire Nuclear

K. Crane, Senior Licensing Specialist

D. Englebert- Welding

C. Fletcher, Corporate Compliance

J. Gabbert, Chemistry Manager

J. Glenn, Organizational Effectiveness Manager

M. Hatley- Site SG Program

M. Isenhour, Engineering Supervisor

B. Kandell- BACC

M. Kelly, Outage and Scheduling Manager

A. Khan, Engineering Supervisor

N. Kunkel, Engineering Manager

S. Mooneyhan, Radiation Protection Manager

C. Morris, Station Manager

B. Richards- Licensing

J. Robertson, Regulatory Affairs Manager

P. Schuerger, Training Manager

S. Snider, Engineering Manager

M. Wilder, Engineering Supervisor

LIST OF REPORT ITEMS

Opened and Closed

05000369, 370/2014005-01 NCV Failure to Adequately Control Transient Combustible Materials and Ignition Sources in Accordance with the Fire Protection Program (Section 1R05)

05000369/2014005-02 NCV Failure to Adequately Implement Containment Closeout Resulting in Loose Debris and Unanalyzed Materials Left in Containment (Section 1R20)

Closed

05000369/2014-002-00 LER Degraded Conditions Due To Rejectable Flaws in 1B and 1C Safety Injection Lines (4OA3)

05000369, 370/2013008-01 URI Fuel Manipulator Crane Digital Modification (Section 4OA5.2)

05000369/2014004-01 URI 1B/1C Reactor Coolant System Loop Safety Injection Piping Flaws (Section 4OA5.3)

TI 2515/190 TI Inspection of the Licensees Proposed Interim Actions as a Result of the Near-Term Task Force Recommendation 2.1 Flooding Reevaluation (Section 4OA5.4)