IR 05000331/2014005

IR 05000331/2014005; 10/01/2014 - 12/31/2014; Duane Arnold Energy Center; Inservice Inspection Activities; Operability Determinations and Functionality Assessments; Surveillance Testing; Radiological Hazard Assessment and Exposure Controls;

ML15037A046
Person / Time
Site:	Duane Arnold
Issue date:	02/04/2015
From:	Boland A Division Reactor Projects III
To:	Vehec T NextEra Energy Duane Arnold
References
EA-15-014 IR 2014005
Download: ML15037A046 (79)

Text

UNITED STATES bruary 4, 2015

SUBJECT:

DUANE ARNOLD ENERGY CENTER - NRC INTEGRATED INSPECTION REPORT 05000331/2014005

Dear Mr. Vehec:

On December 31, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Duane Arnold Energy Center. The enclosed report documents the results of this inspection, which were discussed on January 8, 2015, with you and other members of your staff.

Based on the results of this inspection, eight NRC-identified findings and one self-revealed finding, all of very low safety significance, were identified. These findings involved violations of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your corrective action program, the NRC is treating these issues as non-cited violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy.

Additionally, a licensee-identified violation is listed in Section 4OA7 of this report.

A violation involving the failure to set secondary containment during operations with the potential to drain the reactor vessel (OPDRV) was identified. Specifically, on October 12, 2014, and October 13, 2014, Duane Arnold Energy Center performed a local power range monitor replacement without setting secondary containment, which was a violation of Technical Specification (TS) 3.6.4.1. The NRC issued EGM 11-003, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential for Draining the Reactor Vessel, Revision 2, on December 13, 2013, allowing for the exercise of enforcement discretion for such OPDRV-related TS violations, when certain criteria were met. The NRC concluded that Duane Arnold Energy Center met these criteria. Because the violation was identified during the discretion period described in EGM 11-003, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforcement action for this violation, subject to a timely licensee amendment request being submitted. If you contest the subject or severity of any NCV, you should provide a response within 30 days of the date of this inspection report, with the bases for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Duane Arnold Energy Center. In addition, if you disagree with the cross-cutting aspect assigned to any finding 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 III, and the NRC Resident Inspector at the Duane Arnold Energy Center.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA John B. Giessner Acting for/

Anne T. Boland, Director Division of Reactor Projects Docket No. 50-331 License No. DPR-49

Enclosure:

Inspection Report 05000331/2014005 w/Attachment: Supplemental Information

REGION III==

Docket No: 50-331 License No: DPR-49 Report No: 05000331/2014005 Licensee: NextEra Energy Duane Arnold, LLC Facility: Duane Arnold Energy Center Location: Palo, IA Dates: October 1 through December 31, 2014 Inspectors: L. Haeg, Senior Resident Inspector J. Steffes, Resident Inspector R. Murray, Senior Resident Inspector, Quad Cities C. Phillips, Project Engineer G. ODwyer, Reactor Engineer M. Jeffers, Reactor Inspector J. Gilliam, Senior Reactor Inspector M. Phalen, Senior Health Physicist V. Myers, Health Physicist R. Ng, Project Engineer J. Beavers, Emergency Preparedness Inspector T. Bilik, Senior Reactor Inspector N. Feliz-Adorno, Senior Reactor Inspector S. Smith, Commercial Grade Dedication (CGD) Inspection Team Lead, NRR J. Jacobson, CGD Team, NRR J. Jimenez, CGD Team, NRR J. Heath, CGD Team, NRR A. Dahbur, CGD Team, Region III Approved by: Anne T. Boland, Director Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

Inspection Report 05000331/2014005; 10/01/2014 - 12/31/2014; Duane Arnold Energy Center;

Inservice Inspection Activities; Operability Determinations and Functionality Assessments;

Surveillance Testing; Radiological Hazard Assessment and Exposure Controls; In-Plant Airborne Radioactivity Control and Mitigation; Semi-Annual Trend Review; and Other Activities.

This report covers a three-month period of inspection by resident inspectors, announced baseline inspections by regional inspectors, and a pilot inspection performed by headquarters and regional inspectors. Eight Green findings were identified by the inspectors and one Green finding was self-revealed. The findings were considered non-cited violations (NCVs) of U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of inspection findings is 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 dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas effective dated December 4, 2014. 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 5, dated February, 201

NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

A finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Section 50.55a, Codes and Standards, was identified by the inspectors for the failure to reconcile the construction code and owners requirements when replacing rod hangers associated with the high pressure coolant injection (HPCI) system. The licensee subsequently performed a code reconciliation and concluded the applicable construction code requirements were met.

The licensee captured this issue in its Corrective Action Program as condition report 01999594.

The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of HPCI to respond to initiating events to prevent undesirable consequences.

Specifically, the failure to reconcile the construction code and owners requirements when replacing HPCI support rod hangers reduced confidence in the systems capability to meet its mitigating function consistent with its design basis. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality. This finding had a cross-cutting aspect of procedure adherence in the area of Human Performance because the licensee failed to follow American Society for Mechanical EngineersSection XI, Administrative Manual for Repair, Replacement, and Modification. [H.8] (Section 1R08.1.b.(1)

Green.

A finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments. Specifically, on May 8, 2014, the licensee failed to properly evaluate functionality of a leaking pipe snubber associated with the A core spray subsystem, the resultant operability impact on the Technical Specification affected systems, and the extent of condition. The licensee entered the inspectors concerns into the Corrective Action Program as condition report 02003867 and 02010686. Corrective actions included coaching/training of licensed operators during requalification training and management review committee members, and changes to applicable snubber program procedures.

The performance deficiency was determined to be more than minor because it impacted the Mitigating System cornerstone attribute of equipment performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Because the finding did not involve the total loss of any safety function, the finding screened as very low safety significance (Green). This finding was associated with the cross-cutting aspect of consistent process in the area of Human Performance, because the licensees inconsistent application of the systematic operability/functionality determination process to evaluate the leaking snubber led to prolonged exposure of the extent of cause that affected several safety-related systems. [H.13] (Section 1R15.b.2)

Green.

A finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, when new degraded or non-conforming conditions adverse to quality were identified. Specifically, the licensee failed to evaluate operability and the acceptability for continued operation when an extent of condition review identified several safety-related time delay relays installed beyond the vendor recommended design life. The licensee documented the inspectors concerns in condition report 02015742. The affected relays were immediately declared operable but non-conforming, and a prompt operability determination and apparent cause evaluation to determine corrective actions were in progress at the end of the inspection period.

The performance deficiency was determined to be more than minor because, if left uncorrected, failing to properly assess the operability of degraded or non-conforming conditions and evaluating the necessity for compensatory measures would have the potential to lead to a more significant safety concern. Because the finding was a qualification deficiency confirmed not to result in loss of operability, the finding screened as very low safety significance (Green). This finding was associated with the cross-cutting aspect of identification in the area of Problem Identification & Resolution because the licensee did not identify or capture the extent of the relay aging condition within the corrective action program to ensure that new conditions adverse to quality were properly screened for significance and potential operability impacts. [P.1] (Section 4OA2.3.b)

Green.

The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Part 50,

Appendix B, Criterion III, "Design Control," for the licensee's failure to verify commercial grade circuit breakers were suitable for use in safety-related applications. Specifically, the licensee failed to verify, either through seismic testing or justification, that the circuit breakers being dedicated on purchase order 02309726 would be able to perform their intended safety function during a seismic event. The licensee entered this finding into their Corrective Action Program as condition report 01986727 and 01987616. An extent of condition review was performed and concluded that these circuit breakers were not yet installed at Duane Arnold and a seismic test would be performed on these types of breakers prior to installation.

The performance deficiency was determined to be more than minor because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Because the finding did not represent an actual loss of function (circuit breakers were not currently installed), the finding screened as very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of licensee's current performance. (Section 4OA5.1)

Cornerstone: Barrier Integrity

Green.

A finding of very low safety significance and an associated non-cited violation of Technical Specification 5.4.1.a, Procedures, was identified by the inspectors for the licensees failure to maintain maintenance planning procedures appropriate for the circumstances that could affect performance of safety related equipment. Specifically, procedure MA-AA-203-1001, Work Order Planning, did not ensure that maintenance activities performed on secondary containment components between surveillance testing intervals (2012 and 2014) was properly evaluated for the potential for preconditioning.

The licensee entered the inspectors concerns into the Corrective Action Program as condition report 02008529. Corrective actions included the performance of a condition evaluation to evaluate the work that had been done over the previous cycle for preconditioning and an apparent cause evaluation for the work planning procedural gap with respect to preconditioning and its possible impact on work activities.

The performance deficiency was determined to be more than minor because the finding impacted the Barrier Integrity cornerstone attribute of procedural quality, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents and events. Furthermore, the finding was determined to be more than minor because if left uncorrected, failing to properly and consistently evaluate the potential for unacceptable preconditioning would have the potential to lead to a more significant safety concern. The inspectors applied IMC 0609,

Attachment 4, Initial Characterization of Findings, issued June 19, 2012, to this finding.

The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A,

The Significance Determination Process for Findings At-Power, June 19, 2012. The inspectors answered No to questions C.1 and C.2 in Exhibit 3, Barrier Integrity Screening Questions. Therefore, the finding was screened as very low safety significance (Green). This finding was associated with the cross-cutting aspect of work management in the area of Human Performance because the licensees work order planning process was not appropriate for the circumstances to evaluate the impact of maintenance activities on Technical Specification equipment and surveillance test results. [H.5] (Section 1R15.b.1)

Green.

The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Part 50,

Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish an adequate procedure for an activity affecting quality for a system penetrating the primary containment pressure boundary. Specifically, Surveillance Test Procedure STP 3.6.1.1-09, Containment Isolation Valve Leak Tightness Test - Type C Penetrations - TIP [traversing in-core probe] Valves, Revision 4, failed to include leak rate testing instructions for all of the fittings inboard of the outboard TIP valves tested, which constituted part of the primary containment pressure boundary. The licensee entered the issue in their Corrective Action Program as condition report 02003580. As part of their corrective actions, the licensee re-performed a local leakage rate test to verify the fittings were leak tight.

The inspectors determined that the finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of procedure quality and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. As it related to the finding, procedure STP 3.6.1.1-09 lacked adequate instructions to ensure no leakage of a system penetrating the primary containment pressure boundary. The finding was of very low safety significance (Green) because it did not represent an actual open pathway of reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding was associated with the cross-cutting aspect of resources in the area of Human Performance because STP 3.6.1.1-09 did not include testing of the fittings inboard of the outboard TIP valve as required. [H.1] (Section 1R22.b)

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

Green.

A finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion IX,

Control of Special Processes, was identified by the inspectors for the failure to properly qualify nondestructive testing procedures in accordance with applicable codes.

Specifically, liquid penetrant testing procedures were not qualified for their full applicability temperature ranges in accordance with American Society for Mechanical Engineers (ASME) Code,Section V, Nondestructive Examination. The licensee entered this issue into the Corrective Action Program as condition report 01950601 and 01999596. As an immediate corrective action, the licensee reviewed completed liquid penetrant examination records and did not find an example where the procedures were implemented at the unqualified temperature range.

The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern.

Specifically, since the liquid penetrant testing procedures were not qualified for their full applicability temperature ranges, liquid penetrant examinations were not assured to detect flaws in the unqualified temperature ranges. As a consequence, the potential would exist for a rejectable flaw to go undetected affecting the operability of the affected system. This finding affected the Initiating Event, Mitigating System, and Barrier Integrity cornerstones. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the inadequate qualifications were performed more than three years ago and was not confirmed to reflect current performance. (Section 1R08.1.b.(2))

Cornerstone: Occupational Radiation Safety

Green.

A finding of very low safety significance and an associated non-cited violation of Technical Specification 5.7.1.e was identified by the inspectors following entry into the fuel pool heat exchanger room which was a high radiation area (HRA). The inspectors determined that the licensee failed to determine the radiological conditions in the HRA in accordance with the Technical Specifications and plant procedures to ensure the workers were accurately briefed on the current conditions prior to entry. As a result, an individual was permitted entry into areas with greater than expected dose rates. This issue was entered into the licensees Corrective Action Program as condition report 02000258. The licensee subsequently performed a follow-up survey of the HRA and coached the individual that performed the brief.

The performance deficiency was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, worker entry into HRAs without knowledge of the radiological conditions placed them at increased risk for unnecessary radiation exposure. The finding was determined to be of very low safety significance (Green)because the performance deficiency was not an as-low-as-reasonably-achievable (ALARA) planning issue; there was neither an overexposure nor a substantial potential for an overexposure; and the licensees ability to assess dose was not compromised.

The finding was associated with the cross-cutting aspect of challenge the unknown in the area of Human Performance because the licensee failed to challenge the adequacy of the January 19, 2014, radiological survey as the fuel pool cooling heat exchanger room contained equipment that continuously transported radioactive liquid and was subject to changing radiological conditions. [H.11] (Section 2RS1.2.b)

Green.

A finding of very-low-safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Section 20.1701 was self-revealed during work activities associated with the failure to implement effective radiological engineering controls during reactor pressure vessel (RPV) disassembly that resulted in personal contaminations and unplanned and unintended radiological intakes to workers.

Specifically, on October 5, 2014, several individuals working on the refuel floor were contaminated and several received small intakes of radioactive material while venting the RPV head. The licensee entered the issue into the Corrective Action Program as condition report 01996216. Corrective actions included revising applicable procedures for RPV flood-up with the RPV vented to atmosphere on the refuel floor.

The finding was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the failure to implement effective radiological engineering controls during RPV disassembly resulted in personal contaminations and low dose intakes to several workers. The inspectors also concluded that the radiological hazards had the potential to result in higher exposures to the individuals had the circumstances been slightly altered. The finding was determined to be of very-low-safety significance because it was not an ALARA planning issue; there was neither overexposure nor a substantial potential for an overexposure; and the licensees ability to assess dose was not compromised. This finding was associated with the cross-cutting aspect of operating experience in the area of Problem Identification and

Resolution because the licensee did not systematically implement relevant external operating experience in a timely manner. [P.5] (Section 2RS3.2.b)

Licensee-Identified Violations

A violation of very low safety or security significance or Severity Level IV that was identified by the licensee has been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees Corrective Action Program. These violations and the corrective action program tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Duane Arnold Energy Center (DAEC) operated at full power until October 4, 2014, when the main generator was removed from service to perform a planned refueling outage (RFO). The refueling outage continued through November 26, 2014, when the main generator was connected to the electrical grid. On December 2, 2014, while at 55 percent reactor power, a recirculation system runback occurred and power was reduced to approximately 38 percent due to a failure of an A feedwater system inverter power supply. The station then lowered reactor power to 35 percent using control rods to establish margin to the power-to-flow map buffer zone.

Following the runback and subsequent repairs, the station operated at approximately 53 percent reactor power until the end of the inspection period due to ongoing troubleshooting of elevated vibrations associated with the B reactor feedwater pump and motor.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

1R01 Adverse Weather Protection

.1 Winter Seasonal Readiness Preparations

a. Inspection Scope

The inspectors conducted a review of the licensees preparations for winter conditions to verify that the plants design features and implementation of procedures were sufficient to protect mitigating systems from the effects of adverse weather. Documentation for selected risk-significant systems was reviewed to ensure that these systems would remain functional when challenged by inclement weather. During the inspection, the inspectors focused on plant specific design features and the licensees procedures used to mitigate or respond to adverse weather conditions. Additionally, the inspectors reviewed the Updated Final Safety Analysis Report (UFSAR) and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant specific procedures. Cold weather protection, such as heat tracing and area heaters, was verified to be in operation where applicable. The inspectors also reviewed corrective action program (CAP) items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their CAP in accordance with station corrective action procedures.

Documents reviewed are listed in the Attachment to this report. The inspectors reviews focused specifically on the following plant systems due to their risk significance or susceptibility to cold weather issues:

• Intake, pump house and recirculation motor-generator set room ventilation systems.

This inspection constituted one winter seasonal readiness preparations sample as defined in Inspection Procedure (IP) 71111.01-05.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspectors a performed partial walkdown of the following risk-significant systems:

• Fuel pool cooling and skimmer surge tank level control during period of alternate decay heat removal.

The inspectors selected these systems based on their risk significance relative to the Reactor Safety cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, UFSAR, Technical Specification (TS) requirements, outstanding work orders (WOs), condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the CAP with the appropriate significance characterization. Documents reviewed are listed in the to this report.

These activities constituted one quarterly partial system walkdown sample as defined in IP 71111.04-05.

b. Findings

No findings were identified.

.2 Correction to NRC Inspection Report 05000331/2014002

In NRC Inspection Report 05000331/2014002, the inspectors inadvertently documented a semi-annual complete system walkdown of the A low pressure emergency core cooling system. That inspection sample should have been documented as a quarterly partial system walkdown instead. As such, that previous inspection constituted one quarterly partial system walkdown sample as defined in IP 71111.04-05.

.3 Semi-Annual Complete System Walkdown

a. Inspection Scope

During the weeks of November 30, 2014, and December 7, 2014, the inspectors performed a complete system alignment inspection of the 125 and 250 volt direct current (Vdc) battery systems to verify the functional capability of the systems. These systems were selected because they were considered both safety significant and risk significant in the licensees probabilistic risk assessment. The inspectors walked down the systems to review electrical equipment lineups; electrical power availability; component labeling; component and equipment cooling; hangers and supports; operability of support systems; and to ensure that ancillary equipment or debris did not interfere with equipment operation. A review of a sample of past and outstanding WOs was performed to determine whether any deficiencies significantly affected the systems functions. In addition, the inspectors reviewed the CAP database to ensure that system equipment alignment problems were being identified and appropriately resolved.

Documents reviewed are listed in the Attachment to this report.

These inspections constituted one semi-annual complete system walkdown sample as defined in IP 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Routine Resident Inspector Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

• Off Gas Retention Building;

• Control Building Elevation 800;

• Low Level Radwaste Processing and Storage Facility; Zones 21-A through 21-K; and

• Low Level Radwaste Processing and Storage Facility; Zones 21-L through 21-U.

The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and implemented adequate compensatory measures for out-of-service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event.

Using the documents listed above and in the Attachment to this report, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees CAP.

Documents reviewed are listed in the Attachment to this report.

These activities constituted four routine resident inspector tour samples as defined in IP 71111.05-05.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

From October 14, 2014, through October 17, 2014, the inspectors conducted a review of the implementation of the licensees inservice inspection (ISI) program for monitoring degradation of the reactor coolant system, risk significant piping and components, and containment systems.

The ISI activities described in Sections 1R08.1 and 1R08.5 below constituted one inspection sample as defined in IP 71111.08-05.

.0 Piping Systems In-Service Inspection

a. Inspection Scope

The inspectors either observed or reviewed the following non-destructive examinations mandated by the American Society of Mechanical Engineers (ASME)Section XI Code to evaluate compliance with the ASME Code Section XI and Section V requirements, and if any indications and defects were detected, to determine if these were dispositioned in accordance with the ASME Code or a NRC-approved alternative requirement.

• Ultrasonic (UT) examination of pipe - 90 degree long radius elbow weld in the feedwater system, weld FWC-J006;

• UT of reducer-to-pipe weld in the control rod drive return system, weld CRA-F004;

• UT of safe-end-to-nozzle weld in the control rod drive return system, weld CRA-F002;

• Dye Penetrant examination of the reactor coolant system hanger lug weld, 0200-W-12RCA-K009;

• Magnetic particle examination of reactor vessel (stud holes 40-60), HCC-C001;

• Visual (VT)-3, examination of main-steam snubber, MSA-K011B; and

• VT-3 of mechanical attachment, GBC-4-SR-73.

No examination records completed during the previous outage with relevant/recordable conditions/indications accepted for continued service were identified. Therefore, no NRC review was completed for this inspection procedure attribute.

The inspectors reviewed records for the following pressure boundary weld repairs completed for risk significant systems during the last outage to determine if the licensee applied the pre-service non-destructive examinations and acceptance criteria required by the Construction Code, and/or the NRC-approved Code relief request. Additionally, the inspectors reviewed the welding procedure specifications and supporting weld procedure qualification records to determine whether the weld procedures were qualified in accordance with the requirements of the Construction Code and the ASME Code,Section IX.

• Modify pipe support EBB-14-H-14 (WO 1287016);

• Core spray min-flow wall below 87.5 percent min wall (WO 40052460);

• Replace 3 pipes between 12 HLE-14 and MO1935 (WO 40187789); and

• V24-0046: Perform weld repair disc and hinge (WO 40136087).

b. Findings

(1) Construction Code Used for a Replacement Activity Was Not Reconciled With the Owners Requirements

Introduction:

A finding of very low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulation (CFR), Section 50.55a, Codes and Standards, was identified by the inspectors for the licensees failure to reconcile the construction code and owners requirements during a replacement activity. Specifically, the licensee replaced rod hangers associated with high pressure coolant injection (HPCI) system supports using a different construction code from the original without performing a reconciliation of the code requirements.

Description:

During the 2012 RFO 23, the licensee replaced two rod hangers associated with HPCI support EBB-14-H-14. The piping and supports were originally designed and constructed to the 1969 Edition of B31.7, Nuclear Power Piping Code. Alternatively, the replacement rod hangers were purchased to meet the ASME Code,Section III, Rules for Construction of Nuclear Power Plant Components. The licensee concluded that the replacement met the code requirements applicable to the system because its procedure, ASME Section XI Administrative Manual for Repair, Replacement, and Modification, Revision 21, stated that all or portions of a later or different construction code may be used for replacement activities.

However, the inspectors noted that the licensees procedure, as well as ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, required a reconciliation of different construction code requirements. Specifically, Section 5.3.13.1 of the procedure stated:

An item to be used for repair/replacement activities shall meet the DAECs Requirements. DAEC Requirements may be revised, provided they are reconciled in accordance with Section 5.2.14 (IWA-4222). Reconciliation documentation shall be prepared using Form NG-013Z.

In addition, Section 5.3.13.2 stated An item to be used for repair/replacement activities shall meet the Construction Code and Section 5.3.13.3 stated:

As an alternative to Section 5.2.13.2 above, the item may meet all or portions of the requirements of different Editions and Addenda of the Construction Code, or Section III when the Construction Code was not Section III, provided the requirements of Section 5.2.14 (IWA-4222) through 5.2.22 (IWA-4226), as applicable, are met.

Construction Code Cases may also be used. Reconciliations required by this procedure shall be documented.

The inspectors noted these procedure requirements were translated from ASME Section XI, Sub-Section IWA-4221, Construction Code, and Owners Requirements.

As a result of failing to meet the ASME code requirements, the licensee captured the concern in its CAP as condition report (CR) 01999594. The licensee subsequently performed a reconciliation to restore Code compliance.

Analysis:

The failure to reconcile the construction code and owners requirements during a replacement activity was contrary to ASME Section XI, Sub-Section IWA-4221, and was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of HPCI to respond to initiating events to prevent undesirable consequences. Specifically, the failure to reconcile the construction code and owners requirements when replacing HPCI support rod hangers reduced confidence in the systems capability to meet its mitigating function consistent with its design basis.

The inspectors determined the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued June 19, 2012. Because the finding impacted the Mitigating Systems cornerstone, the inspectors screened the finding through IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions. The finding screened as very low safety significance (Green)because it did not result in the loss of operability or functionality. Specifically, the licensee performed code reconciliation and concluded the applicable construction code requirements were met.

The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because the licensee did not follow a procedure. Specifically, the licensee failed to follow procedure, ASME Section XI Administrative Manual for Repair, Replacement, and Modification, to reconcile the different code requirements. [H.8]

Enforcement:

Title 10 CFR 50.55a(g)(4) requires, in part, that throughout the service life of a boiling water-cooled nuclear power facility, components (including supports) which are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in ASME Section XI.

The 2001 Edition, through 2003 Addenda of ASME Code Section XI, Subsection IWA-4221, Construction Code and Owners Requirements, stated, in part, an item to be used for repair/replacement activities shall meet the Owners Requirements, which may be revised, provided they are reconciled in accordance with IWA-4222. It also stated, in part, an item to be used for repair/replacement activities may meet all or portions of the requirements of different editions and addenda of the construction code, or Section III when the construction code was not Section III, provided the requirements of IWA-4222 through IWA-4226, as applicable, are met and that the reconciliations required by this article shall be documented.

Contrary to the above, on October 27, 2012, the licensee replaced HPCI support EBB-14-H-14 rod hangers using ASME Section III when the construction code was B31.7 without performing reconciliation in accordance with IWA-4222 to ensure the requirements of IWA-4222 through IWA-4226, as applicable, were met.

The licensee subsequently performed code reconciliation without issue; and therefore, the licensee determined HPCI remained operable.

Because this violation was of very low safety significance and was entered into the licensees CAP as CR 01999594, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000331/2014005-01, Construction Code Used during a Replacement Activity Not Reconciled with the Owners Requirements).

(2) Liquid Penetrant Testing Procedures Were Not Qualified for Their Full Applicability Range

Introduction:

A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors for the failure to properly qualify nondestructive testing procedures in accordance with applicable codes. Specifically, liquid penetrant testing procedures were not qualified for their full applicability temperature ranges in accordance with ASME Code,Section V, Nondestructive Examination.

Description:

While reviewing the licensees procedure qualification/verification for expanded temperature for visible dye penetrant procedures Administrative Control Procedure, (ACP) 1211.3, NDE [non-destructive examination] Procedure for Liquid Penetrant (Visible Dye and Water Washable) PT-1, and ACP 1211.4, NDE Procedure for Visible Dye Penetrant - Expanded Temperature Applications (60 - 350 Degrees F),the inspectors noted that the licensee had failed to properly qualify the procedures for the expanded temperature applications delineated therein.

The inspectors noted that ASME Section V, Article 6, Liquid Penetrant Examination, stated that when it is not practical to conduct a liquid penetrant examination within the temperature range of 50°F [degree Fahrenheit] to 125°F, the examination procedure at the proposed lower or higher temperature range requires qualification of the penetrant materials and processing in accordance with Mandatory Appendix III of this Article.

Appendix III, Paragraph 641.2, Temperature Greater Than 125°F (52°C [degree Celsius]), stated, in part that to qualify a procedure for temperatures above 125°F (52°C), the upper and lower temperature limits shall be established and the procedure qualified at these temperatures. [As an example, to qualify a procedure for the temperature range 126°F (52°C) to 200°F (93°C), the capability of a penetrant to reveal indications on the comparator shall be demonstrated at both temperatures]. As part of the qualification process, the Code required that a control test be established to compare the proposed qualification temperature test such that the indications obtained under the proposed conditions are essentially the same as that obtained during examination at 50°F to 125°F (the Code established/qualified range).

The inspectors noted that the licensee had failed to follow the required methodology as required by the ASME code. Procedure ACP 1211.3 was written to expand the upper range of 125°F as permitted by the code to 150°F. Per the ASME code, this would have required tests to be performed at both 126°F and 150°F to bound the extended range.

The licensee failed to perform the lower bound exam. In addition, the licensee failed to establish the upper temperature limit by allowing the temperature to vary from 150°F to 170°F, rather than holding the temperature constant.

In addition, ACP 1211.4 was written to expand the upper range from 125°F to 350°F as permitted by the code. The licensee performed the qualification at the upper limit of 350°F but failed to correctly establish the lower bound of 126°F, instead selecting a temperature of 170°F.

The licensee captured the concern in its CAP as CR 01999596 (ACP 1211.3) and CR 01999542 (ACP 1211.4). As an immediate corrective action, the licensee reviewed completed liquid penetrant examination records and did not find an example where the procedures were implemented at the unqualified temperature range. The licensee planned to perform demonstrations to bring the procedures back into compliance with the code.

Analysis:

The failure to qualify liquid penetrant testing procedures in accordance with ASME Section V was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, since the liquid penetrant testing procedures were not qualified for their full applicability temperature ranges, liquid penetrant examinations were not assured to detect flaws in the unqualified temperature ranges. As a consequence, the potential would exist for a rejectable flaw to go undetected affecting the operability of the affected system. This finding affected the Initiating Event, Mitigating System, and Barrier Integrity cornerstones.

The inspectors determined the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued June 19, 2012. Because the finding impacted the Initiating Event, Mitigating System, and Barrier cornerstones, the inspectors screened the finding through IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality; thus, the inspectors answered no to all of the screening questions. Specifically, the licensee completed a review of liquid penetrant examination records and did not find an example where the procedures were implemented at the unqualified temperature range.

The inspectors did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the inadequate qualifications were performed more than three years ago.

Enforcement:

10 CFR Part 50, Appendix B, Criterion IX, requires, in part, that measures be established to assure that nondestructive testing are controlled and accomplished using qualified procedures in accordance with applicable codes.

Part T-653, Techniques for Nonstandard Temperatures, of ASME Section V, Article 6, Liquid Penetrant Examination, states that when it is not practical to conduct a liquid penetrant examination within the temperature range of 50°F to 125°F, the examination procedure at the proposed lower or higher temperature range requires qualification of the penetrant materials and processing in accordance with Mandatory Appendix III of this Article.

Appendix III, Paragraph 641.2, Temperature Greater Than 125°F (52°C), states, in part that to qualify a procedure for temperatures above 125°F (52°C), the upper and lower temperature limits shall be established and the procedure qualified at these temperatures. [As an example, to qualify a procedure for the temperature range 126°F (52°C) to 200°F (93°C), the capability of a penetrant to reveal indications on the comparator shall be demonstrated at both temperatures].

Contrary to the above, as of October 16, 2014, the licensee did not properly qualify liquid penetrant testing for procedures ACP 1211.3 and ACP 1211.4 for the temperature ranges of 126 F to 150 F and 126 F to 350 F, respectively. Specifically, the licensee did not qualify the procedures at the lower temperature bound.

The licensee was still completing its planned corrective actions at the end of the inspection. However, the inspectors determined that the continued non-compliance did not present an immediate or ongoing safety concern because the licensee did not identify instances where the procedures were implemented at the unqualified temperature ranges.

Because this violation was of very low safety significance and was entered into the licensees CAP as CR 01999596 (ACP 1211.3) and CR 01999542 (ACP 1211.4), this violation is being treated as a NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000331/2014005-02, Liquid Penetrant Testing Procedures Not Qualified for their Full Applicability Range).

.5 Identification and Resolution of Problems

a. Inspection Scope

The inspectors performed a review of ISI related problems entered into the licensees CAP and conducted interviews with licensee staff to determine if:

• the licensee had established an appropriate threshold for identifying ISI related problems;

• the licensee had performed a root cause evaluation (if applicable) and taken appropriate corrective actions; and

• the licensee had evaluated operating experience and industry generic issues related to ISI and pressure boundary integrity.

The inspectors performed these reviews to evaluate 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 Attachment to this report.

b. Findings

No findings were identified

1R11 Licensed Operator Requalification Program

.1 Resident Inspector Quarterly Review of Licensed Operator Requalification

a. Inspection Scope

On December 9, 2014, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator requalification annual examination to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas of the crew:

• licensed operator performance;

• clarity and formality of communications;

• ability to take timely actions in the conservative direction;

• prioritization, interpretation, and verification of annunciator alarms;

• correct use and implementation of abnormal and emergency procedures;

• control board manipulations;

• oversight and direction from supervisors; and

• ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one resident inspector quarterly review of licensed operator requalification sample as defined in IP 71111.11-05.

b. Findings

No findings were identified.

.2 Resident Inspector Quarterly Observation of Heightened Activity or Risk

a. Inspection Scope

On October 3-4, 2014, the inspectors observed control room operators during a planned shut down for RFO 24. This was an activity that required heightened awareness or was related to increased risk. The inspectors evaluated the following areas of the crew:

• licensed operator performance;

• clarity and formality of communications;

• ability to take timely actions in the conservative direction;

• prioritization, interpretation, and verification of annunciator alarms;

• correct use and implementation of procedures;

• control board manipulations;

• oversight and direction from supervisors; and

• ability to identify and implement appropriate TS actions.

The performance in these areas was compared to pre-established operator action expectations, procedural compliance and task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one resident inspector quarterly observation of heightened activity or risk sample as defined in IP 71111.11-05.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

.1 Routine Quarterly Evaluations

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk significant systems:

• Common-cause snubber leakage issue affecting various systems; and

• Secondary containment margin issues.

The inspectors reviewed events such as where ineffective equipment maintenance had resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• implementing appropriate work practices;

• identifying and addressing common cause failures;

• scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

• characterizing system reliability issues for performance;

• charging unavailability for performance;

• trending key parameters for condition monitoring;

• ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and

• verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2), or appropriate and adequate goals and corrective actions for systems classified as (a)(1).

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

This inspection constituted two routine quarterly evaluation samples as defined in IP 71111.12-05.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

• Shutdown risk during Class 1 leak test and scram time testing; and

• Rod drift while inserting fuel assembly in core location 15-08.

These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.

Documents reviewed are listed in the Attachment to this report.

These inspections constituted two maintenance risk assessments and emergent work control samples as defined in IP 71111.13-05.

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors reviewed the following issues:

• Operations with the potential to drain the reactor vessel during RFO 24;

• Pipe snubber functionality impact on TS systems;

• Torus internal coating delamination operability determination to support transition to Modes 1-3;

• Main steam isolation valve limit switch cable jacketing found brittle;

• A source range monitor found not meeting surveillance requirements; and

• Potential preconditioning of secondary containment.

The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and UFSAR to the licensees evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the to this report.

This inspection constituted six operability evaluation samples as defined in IP 71111.15-05.

b. Findings

(1) Failure to Evaluate Maintenance Activities for Preconditioning

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated NCV of TS 5.4.1.a, Procedures, for the licensees failure to maintain maintenance procedures appropriate for the circumstances that could affect performance of safety related equipment. Specifically, procedure MA-AA-203-1001, Work Order Planning, Revision 2, did not ensure that maintenance activities performed on secondary containment components between surveillance testing intervals (2012 and 2014) was properly evaluated for the potential for preconditioning.

Description:

Technical Specifiction Surveillance Requirement 3.6.4.1.3 demonstrates that each Standby Gas Treatment (SBGT) subsystem could maintain greater than or equal to 0.25 inches of water column vacuum in the secondary containment at a flow rate of less than or equal to 4000 cubic feet per minute (cfm). In November of 2012, licensee personnel performed surveillance test procedure, STP 3.6.4.1-01B, Secondary Containment Integrity Using SBGT Train B, to satisfy Surveillance Requirement 3.6.4.1.3, but failed to meet the above acceptance criteria. The licensee performed a root cause evaluation (RCE) and determined that a previous damper maintenance was the cause of the failed surveillance as documented in CRs 01826460 and 01834595. As part of the correction actions, the licensee made flow rate adjustments to both trains of SBGT. Previous flow rates were set at 3800 cfm for both trains, but after adjustment, both A and B trains were set at 4000 cfm and 3950 cfm respectively as documented in CR 01826680. The secondary containment integrity test using separate SBGT subsystems was then performed at the increased flow rates and both SBGT subsystems were able to independently achieve the required vacuum of 0.28 and 0.25 inches of water column, respectively.

On January 28, 2013, the licensee posed a question in CR 01842830 as to why the lack of secondary containment margin was not being pursued with respect to the as-left vacuum readings. This CR resulted in the station making secondary containment margin recovery an equipment reliability issue for resolution throughout the next operating cycle. The site prioritized secondary containment work to improve margin of the as-left vacuum readings from the 2012 tests. However, the licensee did not re-perform the surveillance test after the secondary containment work and the effect on the secondary containment from this maintenance was therefore unknown.

On November 15, 2014, DAEC personnel performed surveillance test STP 3.6.4.1-01A, Secondary Containment Integrity Using SBGT Train A, and achieved 0.25 inches of vacuum water gauge. The inspectors noted that even though the A train of SBGT was able to perform its required safety-related function, there appeared to be a decrease in system performance. Given the surveillance test results, the inspectors questioned what work had been done over the previous cycle and why performance had decreased since the work was performed to improve margin. Since post-maintenance testing for the secondary containment margin recovery work did not include re-performance of either STP 3.6.4.1-01A or STP 3.6.4.1-01B (or a commensurate post maintenance test), the inspectors questioned what process the licensee used to ensure that work performed had not unintentionally introduced deficiencies which could have questioned the ability for secondary containment to perform its safety-related function. Furthermore, the inspectors questioned what process and procedures the licensee used to ensure that corrective and preventative maintenance that was performed did not precondition the secondary containment system for testing.

The inspectors reviewed DAECs work management process procedures and noted that procedure MA-AA-203-1001, Work Order Planning, Revision 2, only contained a brief note to consider the potential for preconditioning. In comparison, the inspectors noted that procedure ER-AA-113-1000, Inservice Testing Procedure, Revision 0, contained a structured and systematic approach to evaluate for potential preconditioning when scheduling maintenance activities prior to inservice testing. Specifically, the approach taken in ER-AA-113-1000 used the same language and similar structure as NRC Inspection Manual Part 9900: Technical Guidance, Maintenance -

Preconditioning of Structures, Systems, and Components Before Determining Operability. The inspectors also noted that per Technical Specification 5.4.1.a, procedures were required to be established covering the applicable procedures of Regulatory Guide 1.33, Revision 2. Specifically, Section 9.a of Regulatory Guide 1.33 covered, in part, procedures appropriate to the circumstances for pre-planning of maintenance that could affect the performance of safety related equipment.

The licensee entered the inspectors concerns into the CAP as CR 02008529.

Corrective actions included the performance of a condition evaluation to evaluate the work that was performed over the previous cycle for preconditioning. The condition evaluation concluded that the damper work performed during the cycle as a corrective action from the RCE in 2012 did not constitute preconditioning because it was corrective maintenance. The condition evaluation further concluded that other secondary containment work that had been completed throughout the cycle, including work done during the 2014 RFO just before performance of the surveillance test, did not constitute unacceptable preconditioning. However, when answering the question would secondary containment have failed the surveillance without the preconditioning, the licensee noted that there existed a possibility that secondary containment would have failed its surveillance if the preventative maintenances performed on doors would not have been completed.

Analysis:

The inspectors determined that failing to establish a work planning procedure appropriate to the circumstances represented a performance deficiency because it was the result of the licensees failure to meet Technical Specification requirements, and the cause was reasonably within the licensees ability to foresee and correct and should have been prevented.

The performance deficiency was determined to be more than minor and a finding because the finding impacted the Barrier Integrity cornerstone attribute of procedural quality, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents and events. Furthermore, the finding was determined to be more than minor because if left uncorrected, failing to properly and consistently evaluate the potential for acceptable or unacceptable preconditioning would have the potential to lead to a more significant safety concern.

The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, June 19, 2012. The inspectors answered No to questions C.1 and C.2 in Exhibit 3, Barrier Integrity Screening Questions. Therefore, the finding was screened as very low safety significance (Green).

The inspectors determined that the performance deficiency was associated with a cross-cutting aspect of work management in the human performance cross-cutting area, and involving the organization implementing a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees work order planning process was not appropriate for the circumstances to evaluate the impact of maintenance activities on TS equipment and surveillance test results. [H.5]

Enforcement:

Technical Specification 5.4.1.a requires that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February, 1978.

Regulatory Guide 1.33, Revision 2, Appendix A, February, 1978, Section 9.a covers, in part, maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate for the circumstances.

Contrary to the above, between January 28, 2013, and November 21, 2014, the licensee failed to establish maintenance planning procedures appropriate to the circumstances that could affect the performance of safety related equipment. Specifically, procedure MA-AA-203-1001, Work Order Planning, Revision 2, was not appropriate such that after the performance of STP 3.6.4.1-01A and STP 3.6.4.1-01B in 2012, the licensee failed to evaluate maintenance conducted on secondary containment for possible impact to the as-left conditions as well as any possible preconditioning of the as-found conditions during the fall of 2014 RFO.

The licensee was preparing an apparent cause evaluation at the end of the inspection period to review the work order planning procedural gap with respect to preconditioning and its possible impact on work activities and had not determined corrective actions by the end of the inspection period. Because this violation was of very low safety significance and because the issue was entered into the licensees CAP as CR 02008529, consistent with Section 2.3.2 of the Enforcement Policy it is being treated as a NCV. (NCV 05000331/2014004-03, Failure to Evaluate Maintenance Activities for Preconditioning).

(2) Failure to Accomplish Procedure for Leaking Pipe Snubber

Introduction:

The inspectors identified a finding of very low safety significance (Green)and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments. Specifically, on May 8, 2014, the licensee failed to properly evaluate functionality of a leaking pipe snubber associated with the A core spray (CS) subsystem, the resultant operability impact on the affected systems, and the extent of condition.

Description:

On May 8, 2014, a licensee quality control inspector identified a fluid leak associated with the A CS pump torus suction piping snubber HBB-002-SS-008. The quality control inspector noted that the snubber fluid reservoir indicator was low at -2.5 and consulted with the snubber program engineer. The engineer generated CR 01964675 and noted that per vendor recommendations and the Duane Arnold snubber inspection procedures that only indications above -1.5 were considered acceptable. The engineer also noted that according to procedure ACP 1410.14, Guidelines for Snubber Operability, snubber HBB-002-SS-008 not only impacted the A CS subsystem, but also primary containment since the snubber was associated with torus-attached piping. The engineer further noted in the CR description that based on the specific design of the snubber, an indicated reservoir level of -3.996 was required before a bottoming-out condition was reached such that the snubber would be non-functional.

An immediate operability determination was documented as part of CR 01964675. Both the A CS and primary containment were declared operable and fully qualified, and extent of condition was documented as: this type of oil leakage is not common and would be self-revealing during plant walkdowns. All snubbers have oil levels verified during each RFO. The CR was screened by licensees management review committee and corrective actions were approved to add snubber HBB-002-SS-008 to the Fall 2014 RFO 24 work scope, to refill the snubber reservoir, and to perform monthly visual inspections of the snubber to verify no additional adverse loss of fluid until planned repairs. A formal functionality assessment was not performed per column two, Table A, Section 6 of EN-AA-203-1001 to assess compensatory measures, and evaluate and document more rigorous bases for the extent of condition and operability impact on the affected TS systems.

On October 7, 2014, during RFO 24, the licensee began visual inspections of a sampling of snubbers per NS992802, Bergen Paterson Hydraulic Snubbers Visual Inspection, to meet, in part, Technical Requirement Manual surveillance requirement 3.7.2.2. During the course of the visual inspections and subsequent scope expansion, 18 snubbers were found unsatisfactory (i.e. fluid reservoir levels less than - 1.5), requiring removal and as-found bench testing. The licensee surmised that a common cause of the fluid losses was related to rebuild work performed during the 2012 refueling outage. The licensee documented the unusually high number of snubber visual failures in CR 01997737 and performed a common cause evaluation. The licensee determined that maintenance procedure SUPORT-A393-01, Anchor/Darling Hydraulic Shock and Sway Arrestors (Formerly Bergen Paterson), that was used to rebuild the snubbers did not incorporate vendor manual guidance to install thread sealant and apply the appropriate torque to the fittings prior to replacement. All Bergen Paterson snubbers at Duane Arnold were visually inspected during RFO 24 and corrective actions taken to properly install the fluid fittings and revise applicable procedures. The licensee determined that none of the degraded snubber conditions rendered any TS systems inoperable as a result of the leaking reservoir fittings.

The inspectors reviewed CR 01964675 from May of 2014 and questioned why a functionality assessment was not performed for snubber HBB-002-SS-008 per the requirements of EN-AA-203-1001. The inspectors noted that NS992802 specified a reservoir level less than - 1.5 to represent an unsatisfactory condition for applicable snubbers. These snubbers would require removal, bench-testing, and be considered in a degraded or non-conforming status. The inspectors were concerned that the operability determination for the affected TS systems (A CS and primary containment)was not properly informed by a functionality assessment for the snubber condition.

Additionally, the extent of condition statement within CR 01964675 was not adequate considering that the apparent cause of the leaking HBB-002-SS-008 snubber condition was unknown at the time. A simple walkdown of similar snubbers could have identified the common cause conditions much sooner based on readily-available visual inspections of the fluid reservoir indicators.

Analysis:

The inspectors determined that the failure to perform a functionality assessment in accordance with procedure EN-AA-203-1001 for the failure of snubber HBB-002-SS-008 was a performance deficiency. The performance deficiency was determined to be more than minor and a finding because it impacted the Mitigating System cornerstone attribute of equipment performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in not properly assessing functionality of the leaking snubber and not promptly identifying the extent of condition which affected the reliability and capability of several safety-related systems.

The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, to this finding. The inspectors answered No to all questions within Table 3, SDP Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. Per Exhibit 2, Mitigating Systems Screening Questions, the inspectors transitioned to Exhibit 4, External Events Screening Questions, and determined that because the finding did not involve the total loss of any safety function, the finding screened as very low safety significance (Green).

The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the performance deficiency was associated with the cross-cutting aspect of consistent process in the human performance cross-cutting area, and involving individuals using a consistent, systematic approach to make decisions.

Specifically, the licensees inconsistent application of the systematic operability/functionality determination process to evaluate the leaking snubber led to prolonged exposure of the extent of cause that affected several safety-related systems.

[H.13]

Enforcement:

10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions.

Contrary to the above, on May 8, 2014, the licensee failed to properly accomplish procedure EN-AA-302-1001, Operability Determinations/Functionality Assessments, Revision 16, to properly assess functionality, operability, and extent of condition of the affected safety related systems following the discovery of a leaking snubber.

Corrective actions included coaching/training of licensed operators during requalification training and management review committee members, and changes to applicable snubber program procedures to align terminologies between the program and operability/functionality procedures.

Because this violation was of very low safety significance and because the issue was entered into the licensees CAP as CRs 02003867 and 02010686, consistent with Section 2.3.2 of the Enforcement Policy, it is being treated as an NCV.

(NCV 05000331/2014003-04, Failure to Accomplish Procedure for Leaking Pipe Snubber).

1R18 Plant Modifications

a. Inspection Scope

The inspectors reviewed the following modification:

• Replacement of B standby diesel generator (SBDG) electrical cables.

The inspectors reviewed the configuration changes and associated 10 CFR 50.59 safety evaluation screening against the design basis, the UFSAR, and the TS, as applicable, to verify that the modification did not affect the operability or availability of the affected system. The inspectors, as applicable, observed ongoing and completed work activities to ensure that the modifications were installed as directed and consistent with the design control documents; the modifications operated as expected; post-modification testing adequately demonstrated continued system operability, availability, and reliability; and that operation of the modification did not impact the operability of any interfacing systems. As applicable, the inspectors verified that relevant procedure, design, and licensing documents were properly updated. Lastly, the inspectors discussed the plant modification with operations and engineering personnel to ensure that the individuals were aware of how the operation with the plant modification in place could impact overall plant performance. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one plant modification sample as defined in IP 71111.18-05.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• Reactor feedwater pump testing following pump replacement;

• B SBDG testing following maintenance;

• Secondary containment isolation damper testing following maintenance;

• Standby transformer testing following cable replacements;

• Class 1 leak test following outage work affecting primary containment; and

• Scram time testing following control rod drive replacements.

These activities were selected based upon the SSCs ability to impact risk. The inspectors evaluated these activities for the following (as applicable): the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion); and test documentation was properly evaluated. The inspectors evaluated the activities against the TSs, the UFSAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with post-maintenance tests to determine whether the licensee was identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted six post-maintenance testing samples as defined in IP 71111.19-05.

b. Findings

No findings were identified.

1R20 Outage Activities

.1 Refueling Outage Activities

a. Inspection Scope

The inspectors reviewed the Outage Safety Plan (OSP) and contingency plans for RFO 24, conducted October 4, 2014, through November 26, 2014, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. During the RFO, the inspectors observed portions of the shutdown and cooldown processes and monitored licensee controls over the outage activities listed below:

• licensee configuration management, including maintenance of defense-in-depth commensurate with the OSP for key safety functions and compliance with the applicable TS when taking equipment out of service;

• implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing;

• installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error;

• controls over the status and configuration of electrical systems to ensure that TS and OSP requirements were met, and controls over switchyard activities;

• monitoring of decay heat removal processes, systems, and components;

• controls to ensure that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system;

• reactor water inventory controls including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss;

• controls over activities that could affect reactivity;

• maintenance of secondary containment as required by TS;

• licensee fatigue management, as required by 10 CFR 26, Subpart I;

• refueling activities, including fuel handling and sipping to detect fuel assembly leakage;

• startup and ascension to full power operation, tracking of startup prerequisites, walkdown of the drywell (primary containment) to verify that debris had not been left which could block emergency core cooling system suction strainers, and reactor physics testing; and

• licensee identification and resolution of problems related to RFO activities.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted one refueling outage activity sample as defined in IP 71111.20-05.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:

• Core spray check valve operability test (Routine);

• SBDG and standby transformer testing (Routine);

• Main turbine overspeed test (Routine); and

• Traversing in-core probe (TIP) valve leak tightness test (Containment Isolation Valve).

The inspectors observed in-plant activities and reviewed procedures and associated records to determine the following:

• did preconditioning occur;

• the effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing;

• acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis;

• plant equipment calibration was correct, accurate, and properly documented;

• as-left setpoints were within required ranges; and the calibration frequency was in accordance with the TSs, the UFSAR, procedures, and applicable commitments;

• measuring and test equipment calibration was current;

• test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied;

• test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used;

• test data and results were accurate, complete, within limits, and valid;

• test equipment was removed after testing;

• where applicable for inservice testing activities, testing was performed in accordance with the applicable version of Section XI, ASME code, and reference values were consistent with the system design basis;

• where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable;

• where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure;

• where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished;

• prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test;

• equipment was returned to a position or status required to support the performance of its safety functions; and

• all problems identified during the testing were appropriately documented and dispositioned in the CAP.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted three routine surveillance testing samples, and one containment isolation valve sample as defined in IP 71111.22-05.

b. Findings

Inadequate Containment Isolation Valve Leak Tightness Test Procedure

Introduction:

The inspectors identified a finding of very low safety significance (Green)and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to test the inboard fittings of the outboard TIP valves, which constituted part of the primary containment pressure boundary. Specifically, the licensee performed procedure STP 3.6.1.1-09, Containment Isolation Valve Leak Tightness Test - Type C Penetrations - TIP Valves, Revision 4, which failed to include leak rate testing of the fittings inboard of the outboard TIP valves when the fittings were restored after testing of the TIP valves.

Description:

On October 22, 2014, during RFO 24, the licensee executed WO 40254762-1 by performing STP 3.6.1.1-09, Containment Isolation Valve Leak Tightness Test - Type C Penetrations - TIP Valves, Revision 4, for valves V-43-503, 1Q122-BALL, 1Q322-BALL, and 1Q222-BALL. This procedure required the valves to be physically separated from the TIP tubing by breaking open the Swagelok fittings inboard of all four outboard TIP valves.

The inspectors observed the licensee perform the local leak rate test (LLRT) and reassemble the TIP tubing by hand tightening the inboard Swagelok fittings. The inspectors determined the inboard fittings were part of the primary containment pressure boundary and noted procedure STP 3.6.1.1-09 did not require a leak rate test to be performed after the final TIP tubing reassembly. The inspectors questioned how the disconnected inboard fittings would be tested and the licensee responded that the fittings would be tested by an integrated leak rate test. However, an integrated leak rate test was not planned for RFO 24.

The licensee agreed with the inspectors that the fittings were required to be tested. On October 31, 2014, the licensee entered the NRC inspectors questions into the CAP as CR 02003580. The corrective actions included re-performing the local leak rate test to verify the fittings were leak tight and the fittings were acceptable. The licensee also revised STP 3.6.1.1-09 to include the fitting leak tightness as part of the LLRT on the TIP valves, and would determine if it is more effective to add test valves and fittings to the system to permit simplified testing. During the subsequent testing of the A TIP line, the leak rate was found to be 1500 standard cubic centimeters per minute (sccm), which exceeded the acceptance criteria of 550 sccm. The test was re-performed for the A TIP line after the fittings were tightened and a satisfactory leak rate was achieved. All TIP system lines were eventually tested satisfactorily.

Analysis:

The inspectors determined the failure to test the inboard fittings of the outboard TIP valves, which constituted part of the primary containment pressure boundary, was a performance deficiency warranting further review. Specifically, the LLRT procedure STP 3.6.1.1-09, Containment Isolation Valve Leak Tightness Test - Type C Penetrations - TIP Valves, Revision 4, did not include steps to test the inboard fittings of the outboard TIP valves after the valves had been physically disconnected from the piping penetrating the drywell and then reassembled. The performance deficiency was determined to be more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, because it was associated with the Barrier Integrity cornerstone attribute of procedure quality and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events.

The inspectors determined the finding could be evaluated using the significance determination process in accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012. The finding was associated with the reactor containment barrier degraded initiator contributor of the Barrier Integrity cornerstone, in accordance with Table 2, Cornerstones Affected by Degraded Condition or Programmatic Weakness. The inspectors answered No to all questions in Table 3, SDP Appendix Router, and therefore, continued the significance evaluation in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions. Because the inspectors answered No to all reactor containment questions, the finding was determined to be of very low safety significance (Green).

The finding was associated with a cross-cutting aspect of resources in the area of Human Performance. The licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety.

Specifically, STP 3.6.1.1-09 did not include testing of the fittings inboard of the outboard TIP valves as requried. [H.1]

Enforcement:

10 CFR Part 50, Appendix J, Option B,Section I, requires, in part, that Option B, Performance-Based leakage-rate test requirements ensure that

(a) leakage through these containments or systems and components penetrating these containments does not exceed allowable leakage rates specified in the Technical Specifications; and

(b) integrity of the containment structure is maintained during its service life.

10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Contrary to the above, on October 22, 2014, the testing performed per STP 3.6.1.1-09, was not appropriate to the circumstances because the fittings inboard of the outboard TIP valves were not tested after the disconnected valves were reassembled. The licensee failed to establish an adequate procedure for leak testing of a system penetrating the primary containment pressure boundary. The corrective actions included re-performing the local leak rate test on November 13, 2014, to verify the fittings were leak tight and the fittings were acceptable. The licensee also revised STP 3.6.1.1-09 to include the fitting leak tightness as part of the LLRT on the TIP valves, and planned to determine if it is more effective to add test valves and fittings to the system to permit simplified testing. Because this violation was of very low safety significance and because this issue was entered into the licensees CAP as CR 02003580, this violation is being treated as a NCV consistent with Section 2.3.2 of the NRCs Enforcement Policy. (NCV 05000331/2014005-05, Inadequate Containment Isolation Valve Leak Tightness Test Procedure).

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

Regional inspectors performed an in-office review of the latest revisions to the emergency plan, emergency plan annex, and emergency plan implementing procedures as listed in the Attachment to this report.

The licensee transmitted the emergency plan and emergency action level revisions to the NRC pursuant to the requirements of 10 CFR Part 50, Appendix E, Section V, Implementing Procedures. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, this revision is subject to future inspection. The specific documents reviewed during this inspection are listed in the Attachment to this report.

This emergency action level and emergency plan change inspection constituted one sample as defined in IP 71114.04-06.

b. Findings

No findings were identified.

2RS1 Radiological Hazard Assessment and Exposure Controls

This inspection constituted one radiological hazard assessment and exposure control sample as defined in IP 71124.01-05.

.1 Inspection Planning (02.01)

a. Inspection Scope

The inspectors reviewed all licensee performance indicators for the Occupational Exposure cornerstone for follow-up. The inspectors reviewed the results of radiation protection program audits (e.g., licensees quality assurance audits or other independent audits). The inspectors reviewed any reports of operational occurrences related to occupational radiation safety since the last inspection. The inspectors reviewed the results of the audit and operational report reviews to gain insights into overall licensee performance.

b. Findings

No findings were identified.

.2 Radiological Hazard Assessment (02.02)

a. Inspection Scope

The inspectors determined if there had been changes to plant operations since the last inspection that could result in a significant new radiological hazard for onsite workers or members of the public. The inspectors evaluated whether the licensee assessed the potential impact of these changes and had implemented periodic monitoring, as appropriate, to detect and quantify the radiological hazard.

The inspectors reviewed the last two radiological surveys from selected plant areas and evaluated whether the thoroughness and frequency of the surveys where appropriate for the given radiological hazard(s).

The inspectors conducted walkdowns of the facility, including radioactive waste processing, storage, and handling areas to evaluate material conditions and performed independent radiation measurements to verify conditions.

The inspectors selected various radiologically risk-significant work activities that involved exposure to radiation.

For these work activities, the inspectors assessed whether the pre-work surveys performed were appropriate to identify and quantify the radiological hazard and to establish adequate protective measures. The inspectors evaluated the radiological survey program to determine if hazards were properly identified, including the following:

• identification of hot particles;

• the presence of alpha emitters;

• the potential for airborne radioactive materials, including the potential presence of transuranics and/or other hard-to-detect radioactive materials (This evaluation included, as applicable, licensee planned entries into non-routinely entered areas subject to previous contamination from failed fuel);

• the hazards associated with work activities that could suddenly and severely increase radiological conditions and that the licensee had established a means to inform workers of changes that could significantly impact their occupational dose; and

• severe radiation field dose gradients that could result in non-uniform exposures of the body.

The inspectors observed work in potential airborne areas and evaluated whether the air samples were representative of the breathing air zone. The inspectors evaluated whether continuous air monitors were located in areas with low background to minimize false alarms and were representative of actual work areas. The inspectors evaluated the licensees program for monitoring levels of loose surface contamination in areas of the plant with the potential for the contamination to become airborne.

b. Findings

Failure to Determine Dose Rates Prior to Entry into a High Radiation Area

Introduction:

A finding of very low safety significance (Green) and an associated NCV of TS 5.7.1.e was identified by the inspectors following entry into the fuel pool heat exchanger room which was a high radiation area (HRA). The inspectors determined that the licensee failed to characterize the radiological conditions in the HRA in accordance with TS and plant procedures prior to allowing personnel entry into the room.

Consequently, the licensee was unable to accurately brief workers on current radiological conditions in the room prior to their entry. As a result, an individual entered areas with greater than expected dose rates.

Description:

On October 18, 2014, an individual received an HRA brief to enter the fuel pool heat exchanger room which was posted as an HRA. The survey used to brief the individual to the expected radiological conditions had been conducted on January 19, 2014, and indicated that the general area dose rates did not exceed approximately 40 millirem per hour. When exiting the radiologically controlled area after performing a walkdown in the HRA, the individual noted that the maximum dose rate received on his electronic dosimeter was approximately 140 millirem per hour. The individual notified radiation protection of the discrepancy and the licensee subsequently performed a follow-up survey of the HRA. This survey indicated that general area dose rates in the area were elevated compared to the survey completed in January, with maximum general area dose rates of approximately 160 millirem per hour.

Plant procedures allow the use of previously performed surveys to brief HRA entries only if there has been no movement or concentration of radioactive material or changes in plant conditions that could affect radiological conditions in the area. However, the fuel pool cooling heat exchanger room contained equipment that continuously moved processed radioactive liquid. Additionally, since the January survey had been completed, the plant had entered a refueling outage which was a change in plant conditions that could drastically change the amount of radioactivity in the water running through the fuel pool system. The inspectors also concluded that the radiological hazards present in the room were variable and had the potential to result in higher exposures to the individuals had the circumstances been slightly altered.

Analysis:

The inspectors determined that the issue of concern was a performance deficiency because entry was made into an HRA without adequately determining the radiological conditions to allow individuals to be accurately briefed, as required in TS and plant procedures. The inspectors determined that the cause of the performance deficiency was reasonably within the licensees ability to foresee and correct and should have been prevented. The finding was not subject to traditional enforcement since the incident did not have a significant safety consequence, did not impact the NRCs ability to perform its regulatory function, and was not willful.

The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, worker entry into HRAs without knowledge of the radiological conditions placed them at increased risk for unnecessary radiation exposure. Additionally, the inspectors reviewed IMC 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009, and identified Example 6(h) as similar to the performance deficiency. The finding was assessed using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, and was determined to be of very low safety significance (Green) because the performance deficiency was not an As Low-As-Reasonably-Achievable (ALARA) planning issue; there was not an overexposure nor substantial potential for an overexposure; and the licensees ability to assess dose was not compromised.

The inspectors determined that the cause of this issue involved the cross-cutting aspect of challenge the unknown in the Human Performance cross-cutting area. Specifically, the licensee failed to challenge the adequacy of the January radiological survey as the fuel pool cooling heat exchanger room contained equipment that continuously transported radioactive liquid and therefore was subject to changing radiological conditions. [H.11]

Enforcement:

Technical Specification S 5.7.1.e requires, in part, that entry into HRAs be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.

Contrary to the above, on October 18, 2014, an individual was allowed to enter an HRA without being knowledgeable of the radiological conditions because they were not adequately determined prior to the entry.

Since this violation was of very low safety significance and has been entered into the licensees CAP as CR 02000258, this violation is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. The licensee performed follow-up surveys and coached the individual who performed the brief. (NCV 05000331/2014005-06, Failure to Determine Dose Rates Prior to Entry into a High Radiation Area)

.3 Instructions to Workers (02.03)

a. Inspection Scope

The inspectors selected various containers holding non-exempt licensed radioactive materials that could cause unplanned or inadvertent exposure of workers, and assessed whether the containers were labeled and controlled in accordance with 10 CFR 20.1904, Labeling Containers, or met the requirements of 10 CFR 20.1905(g), Exemptions To Labeling Requirements.

The inspectors reviewed radiation work permits used to access high radiation areas and evaluated the specified work control instructions or control barriers.

For these radiation work permits, the inspectors assessed whether allowable stay times or permissible dose (including from the intake of radioactive material) for radiologically significant work under each radiation work permit were clearly identified. The inspectors evaluated whether electronic personal dosimeter alarm set-points were in conformance with survey indications and plant policy.

The inspectors reviewed selected occurrences where a workers electronic personal dosimeter noticeably malfunctioned or alarmed. The inspectors evaluated whether workers responded appropriately to the off-normal condition. The inspectors assessed whether the issue was included in the CAP and dose evaluations were conducted as appropriate.

For work activities that could suddenly and severely increase radiological conditions, the inspectors assessed the licensees means to inform workers of changes that could significantly impact their occupational dose.

b. Findings

No findings were identified.

.4 Contamination and Radioactive Material Control (02.04)

a. Inspection Scope

The inspectors observed locations where the licensee monitored potentially contaminated material leaving the radiological control area and inspected the methods used for control, survey, and release from these areas. The inspectors observed the performance of personnel surveying and releasing of material for unrestricted use and evaluated whether the work was performed in accordance with plant procedures and whether the procedures were sufficient to control the spread of contamination and prevent unintended release of radioactive materials from the site. The inspectors assessed whether the radiation monitoring instrumentation had appropriate sensitivity for the type(s) of radiation present.

The inspectors reviewed the licensees criteria for the survey and release of potentially contaminated material. The inspectors evaluated whether there was guidance on how to respond to an alarm that indicates the presence of licensed radioactive material.

The inspectors reviewed the licensees procedures and records to verify that the radiation detection instrumentation was used at its typical sensitivity level based on appropriate counting parameters. The inspectors assessed whether or not the licensee has established a de facto release limit by altering the instruments typical sensitivity through such methods as raising the energy discriminator level or locating the instrument in a high-radiation background area.

The inspectors selected several sealed sources from the licensees inventory records and assessed whether the sources were accounted for and verified to be intact.

The inspectors evaluated whether any transactions, since the last inspection, involving nationally tracked sources were reported in accordance with 10 CFR 20.2207.

b. Findings

No findings were identified.

.5 Radiological Hazards Control and Work Coverage (02.05)

a. Inspection Scope

The inspectors evaluated ambient radiological conditions (e.g., radiation levels or potential radiation levels) during tours of the facility. The inspectors assessed whether the conditions were consistent with applicable posted surveys, radiation work permits, and worker briefings.

The inspectors evaluated the adequacy of radiological controls, such as required surveys, radiation protection job coverage (including audio and visual surveillance for remote job coverage), and contamination controls. The inspectors evaluated the licensees use of electronic personal dosimeters in high noise areas as high radiation area monitoring devices.

The inspectors assessed whether radiation monitoring devices were placed on the individuals body consistent with licensee procedures. The inspectors assessed whether the dosimeters were placed in the location of highest expected dose or that the licensee properly employed an NRC-approved method of determining effective dose equivalent.

The inspectors reviewed the application of dosimetry to effectively monitor exposure to personnel in high radiation work areas with significant dose rate gradients.

As available, the inspectors reviewed radiation work permits for work within airborne radioactivity areas with the potential for individual worker internal exposures.

For these radiation work permits, the inspectors evaluated airborne radioactive controls and monitoring, including potential for significant airborne levels (e.g., grinding, grit blasting, system breaches, entry into tanks, cubicles, and reactor cavities). The inspectors assessed barrier (e.g., tent or glove box) integrity and temporary high-efficiency particulate air ventilation system operation.

The inspectors examined the licensees physical and programmatic controls for highly activated or contaminated materials (i.e., non-fuel) stored within spent fuel and other storage pools. The inspectors assessed whether appropriate controls (i.e.,

administrative and physical controls) were in place to preclude inadvertent removal of these materials from the pool.

The inspectors examined the posting and physical controls for selected high radiation areas and very-high radiation areas to verify conformance with the occupational performance indicator.

b. Findings

No findings were identified.

.6 Risk-Significant High Radiation Area and Very High Radiation Area Controls (02.06)

a. Inspection Scope

The inspectors discussed with the radiation protection manager the controls and procedures for high risk, high radiation areas and very high radiation areas. The inspectors discussed methods employed by the licensee to provide stricter control of very high radiation area access as specified in 10 CFR 20.1602, Control of Access to Very High Radiation Areas, and Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas of Nuclear Plants. The inspectors assessed whether any changes to licensee procedures substantially reduce the effectiveness and level of worker protection.

The inspectors discussed the controls in place for special areas that had the potential to become very high radiation areas during certain plant operations with first-line health physics supervisors (or equivalent positions having backshift health physics oversight authority). The inspectors assessed whether these plant operations required communication beforehand with the health physics group, so as to allow corresponding timely actions to properly post, control, and monitor the radiation hazards including re-access authorization.

The inspectors evaluated licensee controls for very high radiation areas and areas with the potential to become very-high radiation areas to ensure that an individual was not able to gain unauthorized access to the very-high radiation areas.

b. Findings

No findings were identified.

.7 Radiation Worker Performance (02.07)

a. Inspection Scope

The inspectors observed radiation worker performance with respect to stated radiation protection work requirements. The inspectors assessed whether workers were aware of the radiological conditions in their workplace and the radiation work permit controls/limits in place, and whether their performance reflected the level of radiological hazards present.

The inspectors reviewed radiological problem reports since the last inspection that found the cause of the event to be human performance errors. The inspectors evaluated whether there was an observable pattern traceable to a similar cause. The inspectors assessed whether this perspective matched the corrective action approach taken by the licensee to resolve the reported problems. The inspectors discussed with the radiation protection manager any problems with the corrective actions planned or taken.

b. Findings

No findings were identified.

.8 Radiation Protection Technician Proficiency (02.08)

a. Inspection Scope

The inspectors observed the performance of the radiation protection technicians with respect to all radiation protection work requirements. The inspectors evaluated whether technicians were aware of the radiological conditions in their workplace and the radiation work permit controls/limits, and whether their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.

The inspectors reviewed radiological problem reports since the last inspection that had found the cause of the event to be radiation protection technician error. The inspectors evaluated whether there was an observable pattern traceable to a similar cause. The inspectors assessed whether this perspective matched the corrective action approach taken by the licensee to resolve the reported problems.

b. Findings

No findings were identified.

.9 Problem Identification and Resolution (02.09)

a. Inspection Scope

The inspectors evaluated whether problems associated with radiation monitoring and exposure control were being identified by the licensee at an appropriate threshold and were properly addressed for resolution in the licensees CAP. The inspectors assessed the appropriateness of the corrective actions for a selected sample of problems documented by the licensee that involve radiation monitoring and exposure controls.

The inspectors assessed the licensees process for applying operating experience to their plant.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

This inspection constituted a partial sample as defined in IP 71124.03-05.

.1 Inspection Planning (02.01)

a. Inspection Scope

The inspectors reviewed the plant UFSAR to identify areas of the plant designed as potential airborne radiation areas and any associated ventilation systems or airborne monitoring instrumentation. Instrumentation review included continuous air monitors (continuous air monitors and particulate-iodine-noble-gas-type instruments) used to identify changing airborne radiological conditions such that actions to prevent an overexposure may be taken. The review included an overview of the Respiratory Protection Program and a description of the types of devices used. The inspectors reviewed UFSAR, TSs, and emergency planning documents to identify location and quantity of respiratory protection devices stored for emergency use.

Inspectors reviewed the licensees procedures for maintenance, inspection, and use of respiratory protection equipment including self-contained breathing apparatus, as well as procedures for air quality maintenance.

The inspectors reviewed any reported performance indicators related to unintended dose resulting from intakes of radioactive material.

b. Findings

No findings were identified.

.2 Engineering Controls (02.02)

a. Inspection Scope

The inspectors reviewed the licensees use of permanent and temporary ventilation to determine whether the licensee used ventilation systems as part of its engineering controls (in lieu of respiratory protection devices) to control airborne radioactivity. The inspectors reviewed procedural guidance for use of installed plant systems, such as containment purge, spent fuel pool ventilation, and auxiliary building ventilation, and assessed whether the systems are used, to the extent practicable, during high-risk activities (e.g., using containment purge during cavity flood up).

The inspectors selected installed ventilation systems used to mitigate the potential for airborne radioactivity, and evaluated whether the ventilation airflow capacity, flow path (including the alignment of the suction and discharges), and filter/charcoal unit efficiencies, as appropriate, were consistent with maintaining concentrations of airborne radioactivity in work areas below the concentrations of an airborne area to the extent practicable.

The inspectors selected temporary ventilation system setups (high-efficiency particulate air/charcoal negative pressure units, down draft tables, tents, metal Kelly buildings, and other enclosures) used to support work in contaminated areas. The inspectors assessed whether the use of these systems was consistent with licensee procedural guidance and ALARA concept.

The inspectors reviewed airborne monitoring protocols by selecting installed systems used to monitor and warn of changing airborne concentrations in the plant and evaluated whether the alarms and setpoints were sufficient to prompt licensee/worker action to ensure that doses were maintained within the limits of 10 CFR Part 20 and the ALARA concept.

The inspectors assessed whether the licensee had established trigger points (e.g., the Electric Power Research Institutes Alpha Monitoring Guidelines for Operating Nuclear Power Stations) for evaluating levels of airborne beta-emitting (e.g., plutonium-241) and alpha-emitting radionuclides.

b. Findings

Ineffective Radiological Engineering Controls Resulted in Unplanned and Unintended Radiological Intakes to Workers

Introduction:

A finding of very low-safety significance (Green) and an associated NCV of 10 CFR 20.1701 was self-revealed associated with work activities that included preparing to de-tension the reactor pressure vessel (RPV) head. Specifically, ineffective engineering controls created unexpected increases in refuel floor airborne radiological contamination levels. This resulted in nine workers becoming contaminated and workers receiving up to a nominal 20 millirem of unplanned and unintended internal radiological exposures. The inspectors concluded that the licensee failed to adequately institute process or other engineering controls as required by 10 CFR 20.1701 to minimize workers radiological exposures.

Description:

On October 5, 2014, during RFO 24, the normal RPV disassembly sequence required plant workers to vent the reactor head and raise RPV water (reactor coolant) level up to the RPV head flange prior to reactor head removal. The added water is cold relative to the internal components of the recently shut down reactor. This sequence increased the amount of gasses that must be vented before the reactor head could be safely removed and created several paths that generated airborne radioactivity.

Specifically, reactor flood-up activities displaced the radioactive air volume that was present within the reactor; air movement sweeping past highly contaminated RPV internals increased airborne radioactivity concentrations; some of the radioactive gasses entrained within the reactor coolant were liberated as reactor coolant temperature and pressure decreased; and adding relatively cooler water to the RPV during flood-up flashed some of the water to steam as the cooler water contacted the relatively hotter RPV internals. Consequently, as part of the stations airborne radioactivity management process, the reactor head was initially vented into the drywell equipment drain sump.

Also, reactor head venting and removal activities included removal and re-installation of reactor water level instrumentation important to safe plant operation. Thus, the reactor shut down and disassembly work protocol was closely controlled and sequenced in such a way as to maintain nuclear safety while minimizing the concentration of airborne radioactivity present on the refuel floor. There was extensive industry operating experiences available to the licensee since the early 1990s that identified inadequate reactor flood-up and reactor head venting sequencing activities that resulted in elevated refuel floor radioactive airborne concentrations.

On October 5, 2014, portions of the RPV and internal component temperatures were above 212°F while disconnecting RPV vent flanges and the instrument line flange. The RPV water level was below the level of RPV internal components (i.e., the steam dryer).

This allowed for the water to flash to steam when coming into contact with hot components, vent through the RPV flanges and drive hot radioactive air to the refuel floor.

Station procedures utilized to disassemble the reactor head vent, maintain RPV level indication and RPV flood-up were correctly followed and the sequence of events for the work evolutions were conducted in accordance with the stations outage schedule.

However, by failing to effectively utilize industry operating experiences, the station failed to effectively schedule and sequence reactor disassembly activities in such a way as to minimize the airborne concentrations of radioactive materials on the refuel floor.

Consequently, nine workers became contaminated and workers received up to a nominal 20 millirem of unintended internal radiological exposures.

Analysis:

The inspectors determined that the issue of concern was a performance deficiency because the licensee failed to effectively implement the regulatory required engineering controls that were necessary to demonstrate compliance with the requirements of 10 CFR 20.1701 for the use of engineering controls. The inspectors determined that the cause of the performance deficiency was reasonably within the licensees ability to foresee and correct and should have been prevented.

The finding was not subject to traditional enforcement since the incident did not have a significant safety consequence, did not impact the NRCs ability to perform its regulatory function, and was not willful.

The inspectors determined that that the finding was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, in that the finding impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the failure to implement effective radiological engineering controls during RPV disassembly resulted in nine personal contaminations and low dose intakes to several workers. The inspectors also concluded that the radiological hazards had the potential to result in higher exposures to the individuals had the circumstances been slightly altered. The inspectors also reviewed the guidance in IMC 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009, and did not identify any examples similar to the performance issue. The finding was assessed using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issued October 28, 2011 and the finding was determined to be of very low safety significance because it was not an ALARA planning issue; there was neither overexposure nor a substantial potential for an overexposure; and the licensees ability to assess dose was not compromised.

The licensees evaluation of this issue determined that there was ample industry operating experience available that was not effectively implemented at the station.

Consequently, the inspectors concluded that the cause of the issue involved a cross-cutting aspect of operating experience in the Problem Identification and Resolution area.

Specifically, the licensee did not systematically implement relevant external operating experience in a timely manner. [P.5]

Enforcement:

Title10 CFR 20.1701 states, in part, that the licensee shall use, to the extent practical, process or other engineering controls to control the concentration of radioactive material in air.

Contrary to the above, on October 5, 2014, the licensee failed to adequately implement engineering controls and created unexpected increases in airborne radiological contamination levels. This resulted in nine workers becoming contaminated and some workers receiving up to a nominal 20 millirem of unintended internal radiological exposures.

As part of the corrective actions, the licensee planned to revise applicable procedures for RPV flood-up with the RPV vented to atmosphere on the refuel floor. Additionally, the licensee planned to institute radiological engineering controls for situations that create a treated pathway for ventilation discharges.

Since the failure to adequately implement engineering controls was of very low safety significance and the issue was entered into the licensees CAP as CR 01996216, the violation is being treated as a NCV consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000331/2014005-07, Ineffective Radiological Engineering Controls Resulted in Unplanned and Unintended Radiological Intakes to Workers)

.3 Use of Respiratory Protection Devices (02.03)

a. Inspection Scope

For those situations where it was impractical to employ engineering controls to minimize airborne radioactivity, the inspectors assessed whether the licensee provided respiratory protective devices such that occupational doses were ALARA. The inspectors selected work activities where respiratory protection devices were used to limit the intake of radioactive materials, and assessed whether the licensee performed an evaluation concluding that further engineering controls were not practical and that the use of respirators was ALARA. The inspectors also evaluated whether the licensee had established means (such as routine bioassay) to determine if the level of protection (protection factor) provided by the respiratory protection devices during use was at least as good as that assumed in the licensees work controls and dose assessment.

The inspectors assessed whether respiratory protection devices used to limit the intake of radioactive materials were certified by the National Institute for Occupational Safety and Health/Mine Safety and Health Administration or have been approved by the NRC per 10 CFR 20.1703(b). The inspectors selected work activities where respiratory protection devices were used. The inspectors evaluated whether the devices were used consistent with their National Institute for Occupational Safety and Health/Mine Safety and Health Administration certification or any conditions of their NRC approval.

The inspectors selected several individuals qualified to use respiratory protection devices, and assessed whether they had been deemed fit to use the devices by a physician.

b. Findings

No findings were identified.

2RS4 Occupational Dose Assessment

This inspection constituted one complete sample as defined in IP 71124.04-05.

.1 Inspection Planning (02.01)

a. Inspection Scope

The inspectors reviewed the results of radiation protection program audits related to internal and external dosimetry (e.g., licensee quality assurance audits, self-assessments, or other independent audits) to gain insights into overall licensee performance in the area of dose assessment and focus the inspection activities consistent with the principle of smart sampling.

The inspectors reviewed the most recent National Voluntary Laboratory Accreditation Program accreditation report on the vendors most recent results to determine the status of the contractors accreditation.

A review was conducted of the licensee procedures associated with dosimetry operations, including issuance/use of external dosimetry (routine, multibadging, extremity, neutron, etc.), assessment of internal dose (operation of whole body counter, assignment of dose based on derived air concentration-hours, urinalysis, etc.), and evaluation of and dose assessment for radiological incidents (distributed contamination, hot particles, loss of dosimetry, etc.).

The inspectors evaluated whether the licensee had established procedural requirements for determining when external and internal dosimetry was required.

b. Findings

No findings were identified.

.2 External Dosimetry (02.02)

a. Inspection Scope

The inspectors evaluated whether the licensees dosimetry vendor was National Voluntary Laboratory Accreditation Program accredited and if the approved irradiation test categories for each type of personnel dosimeter used were consistent with the types and energies of the radiation present and the way the dosimeter was being used (e.g., to measure deep dose equivalent, shallow dose equivalent, or lens dose equivalent).

The inspectors evaluated the onsite storage of dosimeters before their issuance, during use, and before processing/reading. The inspectors also reviewed the guidance provided to rad-workers with respect to care and storage of dosimeters.

The inspectors assessed whether non-National Voluntary Laboratory Accreditation Program accredited passive dosimeters (e.g., direct ion storage sight read dosimeters)were used according to licensee procedures to provide for periodic calibration, application of calibration factors, usage, reading (dose assessment) and zeroing.

The inspectors assessed the use of active dosimeters (electronic personal dosimeters)to determine if the licensee used a correction factors to address the response of the electronic personal dosimeter as compared to the passive dosimeter for situations when the electronic personal dosimeter must be used to assign dose. The inspectors also assessed whether the correction factors were based on sound technical principles.

The inspectors reviewed dosimetry occurrence reports or CAP documents for adverse trends related to electronic personal dosimeters, such as interference from electromagnetic frequency, dropping or bumping, failure to hear alarms, etc. The inspectors assessed whether the licensee had identified any trends and implemented appropriate corrective actions.

b. Findings

No findings were identified.

.3 Internal Dosimetry (02.03)

a. Routine Bioassay (In Vivo)

(1) Inspection Scope The inspectors reviewed procedures used to assess the dose from internally deposited nuclides using whole body counting equipment. The inspectors evaluated whether the procedures addressed methods for differentiating between internal and external contamination, the release of contaminated individuals, the route of intake and the assignment of dose.

The inspectors reviewed the whole body count process to determine if the frequency of measurements were consistent with the biological half-life of the nuclides available for intake.

The inspectors reviewed the licensee's evaluation for use of its portal radiation monitors as a passive monitoring system to determine if instrument minimum detectable activities were adequate to determine the potential for internally deposited radionuclides sufficient to prompt additional investigation.

The inspectors selected several whole body counts and evaluated whether the counting system used had sufficient counting time/low background to ensure appropriate sensitivity for the potential radionuclides of interest. The inspectors reviewed the radionuclide library used for the count system to determine its appropriateness. The inspectors evaluated whether any anomalous count peaks/nuclides indicated in each output spectra received appropriate disposition. The inspector's reviewed the licensee's 10 CFR Part 61 data analyses to determine whether the nuclide libraries included appropriate gamma-emitting nuclides. The inspectors evaluated how the licensee accounts for hard-to-detect nuclides in the dose assessment.

(2) Findings No findings were identified.

b. Special Bioassay (In Vitro)

(1) Inspections Scope As available, the inspectors selected internal dose assessments obtained using in vitro monitoring. The inspectors reviewed and assessed the adequacy of the licensees program for in vitro monitoring (i.e., urinalysis and fecal analysis) of radionuclides (tritium, fission products, and activation products), including collection and storage of samples.

The inspectors reviewed the vendor laboratory quality assurance program and assessed whether the laboratory participated in an industry recognized cross-check program including whether out-of-tolerance results were resolved appropriately.

(2) Findings No findings were identified.

c. Internal Dose Assessment - Airborne Monitoring

(1) Inspection Scope The inspectors reviewed the licensee's program for airborne radioactivity assessment and dose assessment, as applicable, based on airborne monitoring and calculations of derived air concentration. The inspectors determined whether flow rates and collection times for air sampling equipment were adequate to allow lower limits of detection to be obtained. The inspectors also reviewed the adequacy of procedural guidance to assess internal dose if respiratory protection was used.

(2) Findings No findings were identified.

d. Internal Dose Assessment - Whole Body Count Analyses

(1) Inspection Scope The inspectors reviewed several dose assessments performed by the licensee using the results of whole body count analyses. The inspectors determined whether affected personnel were properly monitored with calibrated equipment and that internal exposures were assessed consistent with the licensee's procedures.

(2) Findings No findings were identified.

.4 Special Dosimetric Situations (02.04)

a. Declared Pregnant Workers

(1) Inspection Scope The inspectors assessed whether the licensee informed workers, as appropriate, of the risks of radiation exposure to the embryo/fetus, the regulatory aspects of declaring a pregnancy, and the specific process to be used for (voluntarily) declaring a pregnancy.

The inspectors selected individuals who had declared pregnancy during the current assessment period and evaluated whether the licensees radiological monitoring program (internal and external) for declared pregnant workers was technically adequate to assess the dose to the embryo/fetus. The inspectors reviewed exposure results and monitoring controls employed by the licensee and with respect to the requirements of Title 10 CFR Part 20.

(2) Findings No findings were identified.

b. Dosimeter Placement and Assessment of Effective Dose Equivalent for External Exposures

(1) Inspection Scope The inspectors reviewed the licensee's methodology for monitoring external dose in non-uniform radiation fields or where large dose gradients existed. The inspectors evaluated the licensee's criteria for determining when alternate monitoring, such as use of multi-badging, was to be implemented.

The inspectors reviewed dose assessments performed using multi-badging to evaluate whether the assessment was performed consistently with licensee procedures and dosimetric standards.

(2) Findings No findings were identified.

c. Shallow Dose Equivalent

(1) Inspection Scope The inspectors reviewed shallow dose equivalent dose assessments for adequacy. The inspectors evaluated the licensees method (e.g., VARSKIN or similar code) for calculating shallow dose equivalent from distributed skin contamination or discrete radioactive particles.

(2) Findings No findings were identified.

d. Neutron Dose Assessment

(1) Inspection Scope The inspectors evaluated the licensees neutron dosimetry program, including dosimeter types and/or survey instrumentation.

The inspectors reviewed neutron exposure situations (e.g., independent spent fuel storage installation operations or at-power containment entries) and assessed whether

(a) dosimetry and/or instrumentation was appropriate for the expected neutron spectra,

(b) there was sufficient sensitivity for low dose and/or dose rate measurement, and (c)neutron dosimetry was properly calibrated. The inspectors also assessed whether interference by gamma radiation had been accounted for in the calibration and whether time and motion evaluations were representative of actual neutron exposure events, as applicable.

(2) Findings No findings were identified.

e. Assigning Dose of Record

(1) Inspection Scope For the special dosimetric situations reviewed in this section, the inspectors assessed how the licensee assigned dose of record for total effective dose equivalent, shallow dose equivalent, and lens dose equivalent. This included an assessment of external and internal monitoring results, supplementary information on individual exposures (e.g.,

radiation incident investigation reports and skin contamination reports), and radiation surveys and/or air monitoring results when dosimetry was based on these techniques.

(2) Findings No findings were identified.

.5 Problem Identification and Resolution (02.05)

a. Inspection Scope

The inspectors assessed whether problems associated with occupational dose assessment were being identified by the licensee at an appropriate threshold and were properly addressed for resolution in the licensees CAP. The inspectors assessed the appropriateness of the corrective actions for a selected sample of problems documented by the licensee involving occupational dose assessment.

b. Findings

No findings were identified.

OTHER ACTIVITIES

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

4OA1 Performance Indicator Verification

.1 Mitigating Systems Performance Index - Residual Heat Removal System

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance Index (MSPI) - residual heat removal system performance indicator for the period from the fourth quarter 2013, through the third quarter 2014. To determine the accuracy of the performance indicator (PI) data reported during those periods, PI definitions and guidance contained in the Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, was used. The inspectors reviewed the licensees operator narrative logs, issue reports, MSPI derivation reports, event reports and NRC integrated inspection reports for the period of October 2013, through September 2014, to validate the accuracy of the submittals. The inspectors reviewed the MSPI component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one MSPI residual heat removal system sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.2 Safety System Functional Failures

a. Inspection Scope

The inspectors sampled licensee submittals for the Safety System Functional Failures performance indicator for the period from the fourth quarter 2013 through the third quarter 2014. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73," Revision 3, definitions and guidance, were used. The inspectors reviewed the licensees operator narrative logs, operability assessments, maintenance rule records, maintenance work orders, issue reports, event reports and NRC integrated inspection reports for the period of October 2013, through September of 2014, to validate the accuracy of the submittals.

The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one safety system functional failures sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.3 Mitigating Systems Performance IndexCooling Water Systems

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance Index - cooling water systems performance indicator for the period from the fourth quarter 2013 through the third quarter 2014. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, was used. The inspectors reviewed the licensees operator narrative logs, issue reports, MSPI derivation reports, event reports and NRC integrated inspection reports for the period of October of 2013, through September of 2014, to validate the accuracy of the submittals. The inspectors reviewed the MSPI component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the to this report.

This inspection constituted one MSPI cooling water system sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.4 Reactor Coolant System Specific Activity

a. Inspection Scope

The inspectors sampled licensee submittals for the reactor coolant system specific activity PI for DAEC for the period from the third quarter 2013 through the third quarter 2014. The inspectors used PI definitions and guidance contained in the NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, to determine the accuracy of the PI data reported during those periods.

The inspectors reviewed the licensees reactor coolant system chemistry samples, TS requirements, issue reports, event reports and NRC integrated inspection reports to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. In addition to record reviews, the inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one reactor coolant system specific activity sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.5 Occupational Exposure Control Effectiveness

a. Inspection Scope

The inspectors sampled licensee submittals for the Occupational Exposure Control Effectiveness Performance Indicator for the period from the third quarter 2013 through the third quarter 2014. The inspectors used PI definitions and guidance contained in the NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, to determine the accuracy of the PI data reported during those periods.

The inspectors reviewed the licensees assessment of the PI for occupational radiation safety to determine if the indicator related data was adequately assessed and reported.

To assess the adequacy of the licensees PI data collection and analyses, the inspectors discussed with radiation protection staff the scope and breadth of its data review and the results of those reviews. The inspectors independently reviewed electronic personal dosimetry dose rate and accumulated dose alarms and dose reports and the dose assignments for any intakes that occurred during the time period reviewed to determine if there were potentially unrecognized occurrences. The inspectors also conducted walkdowns of numerous locked high and very high radiation area entrances to determine the adequacy of the controls in place for those areas. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one occupational exposure control effectiveness sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.6 Radiological Effluent Technical Specification/Offsite Dose Calculation Manual

Radiological Effluent Occurrences

a. Inspection Scope

The inspectors sampled licensee submittals for the radiological effluent Technical Specification/Offsite Dose Calculation Manual radiological effluent occurrences PI for the period from the third quarter 2013 through the third quarter 2014. The inspectors used PI definitions and guidance contained in the NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, to determine the accuracy of the PI data reported during those periods. The inspectors reviewed the licensees issue report database and selected individual reports generated since this indicator was last reviewed to identify any potential occurrences such as unmonitored, uncontrolled, or improperly calculated effluent releases that may have impacted offsite dose. The inspectors reviewed gaseous effluent summary data and the results of associated offsite dose calculations for selected dates to determine if indicator results were accurately reported. The inspectors also reviewed the licensees methods for quantifying gaseous and liquid effluents and determining effluent dose. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one Radiological Effluent Technical Specification/Offsite Dose Calculation Manual radiological effluent occurrences sample as defined in IP 71151-05.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify they were being entered into the licensees CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Attributes reviewed included: identification of the problem was complete and accurate; timeliness was commensurate with the safety significance; evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent-of-condition reviews, and previous occurrences reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue. Minor issues entered into the licensees CAP as a result of the inspectors observations are included in the Attachment to this report.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees CAP. This review was accomplished through inspection of the stations daily condition report packages.

These daily reviews were performed by procedure as part of the inspectors daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

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 screening discussed in Section 4OA2.2 above, licensee trending efforts, and licensee human performance results. The inspectors review nominally considered the 6-month period from July 2014, through December 2014, although some examples expanded beyond those dates where the scope of the trend warranted.

The review also included issues documented outside the normal CAP in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance audit/surveillance reports, self-assessment reports, and Maintenance Rule assessments. The inspectors compared and contrasted their results with the results contained in the licensees CAP trending reports. Corrective actions associated with a sample of the issues identified in the licensees trending reports were reviewed for adequacy.

This inspection constituted one semi-annual trend inspection sample as defined in IP 71152-05.

b. Findings

Failure to Evaluate Several Safety Related Relays Installed Beyond their Design Life

Introduction:

The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, when new degraded or non-conforming conditions adverse to quality were identified. Specifically, the licensee failed to evaluate operability and the acceptability for continued operation when an extent of condition review identified several safety related time delay relays installed beyond the vendor recommended design life.

Description:

In January 2013, the licensee completed apparent cause evaluation (ACE) 01832534 following the December 13, 2012, test failure of time delay relay (KY-4401) associated with the low pressure coolant injection system. The licensee determined the apparent cause was that the preventative maintenance task frequency did not accurately reflect the replacement frequency recommended by the manufacturer. The failed relay identified during testing was beyond its design life of 10 years from the date of manufacture.

An extent of condition review performed as part of the ACE determined that there were more than 50 similar time delay relays that did not have a preventative maintenance replacement frequency that met the vendor recommendations. Additionally, the licensee noted that there were more than 20 safety-related relays that were currently installed beyond their design life. Corrective actions included planning replacement of relays that were beyond their installed design life and changing the replacement frequency template.

During RFO 24 in the fall of 2014, the licensee generated two CRs for time delay relays associated with the anticipated transient without scram system that were found to be missing from the original extent of condition list. As a result, the licensee recommended replacement of the relays. The inspectors reviewed the CRs, as well as the ACE performed in January 2013, and questioned whether a new CR was generated when the extent of condition identified additional relays beyond their design life. After reviewing the ACE and associated corrective actions, the inspectors were concerned that no new CR(s) had been generated to evaluate operability of the relays identified in the ACE extent of condition, nor were any evaluations performed to support continued operation with the relays until replacement were made. The inspectors identified that the licensee ultimately failed to evaluate the acceptability of continued operation with the relays installed beyond their design life and determined that this was contrary to the requirements of procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 9 (revision in place upon completion of ACE 01832534).

Specifically, Step 1 of Section 4.2, Concern Identification, of EN-AA-203-1001, Revision 9, stated, in part, that A person upon discovering a potential or suspected degraded or nonconforming condition shall ensure the concern is documented in a Condition Report. Since the licensee never documented the extent of condition within the CAP, the shift manager was never afforded the opportunity to review the operability impact on other affected systems.

The licensee documented the inspectors concerns in CR 02015742. The affected relays were immediately declared operable but non-conforming, and a prompt operability determination and apparent cause evaluation were in progress at the end of the inspection period.

Analysis:

The inspectors determined that the failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, when an extent of condition review identified several safety related time delay relays installed beyond the vendor recommended design life, was a performance deficiency. The performance deficiency was determined to be more than minor and a finding because if left uncorrected, failing to properly assess the operability of degraded or non-conforming conditions would have the potential to lead to a more significant safety concern.

Specifically, by not identifying and appropriately evaluating degraded or non-conforming conditions, circumstances could exist that warrant declaring a SSC inoperable, or warrant compensatory measures to maintain or enhance a degraded or non-conforming condition.

The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, to this finding. The inspectors answered No to all questions within Table 3, SDP Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. Per Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding screened as very low safety significance (Green) because the design/qualification deficiency did not affect operability of the mitigating SSCs.

The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the performance deficiency was associated with the cross-cutting aspect of identification in the Problem Identification & Resolution cross-cutting area, and involved the organization implementing a CAP with a low threshold for identifying issues. Specifically, the licensee did not identify or capture the extent of the relay aging condition within the CAP to ensure that new conditions adverse to quality were properly screened for significance and potential operability impacts. [P.1]

Enforcement:

10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented procedures and shall be accomplished in accordance with these procedures.

Contrary to this requirement, on January 24, 2013, the licensee failed to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 9. Specifically, upon determining that an extent of condition existed as related to the apparent cause of the failure of time delay relay KY-4401, the licensee failed to document the condition in the CAP to ensure that the potential degraded/non-conforming conditions were appropriately dispositioned.

Corrective actions included declaring the affected relays operable but non-conforming, and a prompt operability determination and apparent cause evaluation were in progress at the end of the inspection period. Because this violation was of very low safety significance and was entered into the licensees CAP as CR 02015742, the violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.

(NCV 05000331/2014005-08, Failure to Evaluate Several Safety Related Relays Installed Beyond their Design Life)

.4 Selected Issue Follow-Up Inspection: Failure to Meet Design Requirements for Safety

Related Cable Junction Boxes

a. Inspection Scope

The inspectors reviewed RCE 01979556 associated with the licensees discovery that several non-qualified/non-conforming electrical cable splices and terminal strips were installed in the A and B SBDG systems during the 2012 RFO 23. The inspectors determined that the corrective actions were appropriate to address the root and contributing causes of the issue. A licensee-identified violation associated with the issue is discussed in Section 4OA7 of this report.

This review constituted one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

No findings were identified.

.5 Selected Issue Follow-Up Inspection: NRC Cross-Cutting Issue P.2 Adverse Trend

a. Inspection Scope

The inspectors reviewed RCE 01988427 associated with a cross-cutting theme in P.2 (Evaluation) which was initiated by the licensee following the receipt of five NRC findings with a P.2 aspect between the third quarter of 2013 and the third quarter of 2014. The inspectors determined that the corrective actions were reasonable for the identified root and contributing causes; however, no further conclusions could be drawn on effectiveness of the actions at end of the inspection period.

This review constituted one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

No findings were identified.

4OA3 Follow-Up of Events and Notices of Enforcement Discretion

.1 (Closed) Licensee Event Report (LER) 05000331/2013-006-00: Both Doors in

Secondary Containment Airlock Opened Concurrently This event, which occurred on December 18, 2013, involved the simultaneous opening of two doors (Door 225 and 227) while workers were traversing through a secondary containment access airlock. The workers recognized the airlock condition, closed both doors within a few (less than 10) seconds and verified that the doors were latched, and notified the control room. The momentary opening of both doors within the airlock resulted in the station failing to meet TS Surveillance Requirement 3.6.4.1.2 to verify that either door in each secondary containment access opening are closed, and therefore, momentarily rendered secondary containment inoperable per TS Limiting Condition for Operation 3.6.4.1. The licensee investigated the airlock doors and identified that the door 225 magnet needed adjustment. The repairs were made and the secondary containment airlock verification surveillance test was performed satisfactorily to demonstrate functionality of the interlock.

The inspectors reviewed LER 05000331/2013-006-00 against reporting requirements and found no issues. The inspectors also reviewed the licensees assessment of safety consequences in the LER, specifically the basis for not considering the condition a safety system functional failure. The licensees post-loss of coolant accident (LOCA)dose calculation of record did not credit secondary containment integrity for on-site and off-site doses for the first 5 minutes of the event. Therefore, the inspectors determined that it was reasonable to conclude that the simultaneous opening condition of the secondary containment doors was bounded by the existing licensing basis calculation of record. Documents reviewed are listed in the Attachment to this report. This LER is closed.

This inspection constituted one event follow up sample as defined in IP 71153-05.

.2 (Closed) LER 05000331/2014-002-00: Both Doors in Secondary Containment Airlock

Opened Concurrently This event, which occurred on February 18, 2014, involved the simultaneous opening of two doors (Door 245 and 239A) while workers was traversing through a secondary containment access airlock. The workers recognized the airlock condition, closed both doors within a few (less than 10) seconds and verified that the doors were latched, and notified the control room. The momentary opening of both doors within the airlock resulted in the station failing to meet TS Surveillance Requirement 3.6.4.1.2 to verify that either door in each secondary containment access opening are closed, and therefore, momentarily rendered secondary containment inoperable per TS Limiting Condition for Operation 3.6.4.1. The licensee investigated the airlock doors and identified that the door 245 magnet needed adjustment. The repairs were made and the secondary containment airlock verification surveillance test was performed satisfactorily to demonstrate functionality of the interlock.

The inspectors reviewed LER 05000331/2014-002-00 against reporting requirements and found no issues. The inspectors also reviewed the licensees assessment of safety consequences in the LER, specifically the basis for not considering the condition a safety system functional failure. The licensees post-LOCA dose calculation of record did not credit secondary containment integrity for on-site and off-site doses for the first 5 minutes of the event. Therefore, the inspectors determined that it was reasonable to conclude that the simultaneous opening condition of the secondary containment doors was bounded by the existing licensing basis calculation of record. Documents reviewed are listed in the Attachment to this report. This LER is closed.

This inspection constituted one event follow up sample as defined in IP 71153-05.

.3 (Closed) LER 05000331/2014-003-00: Both Doors in Secondary Containment Airlock

Opened Concurrently This event, which occurred on May 30, 2014, involved the simultaneous opening of two doors (225 and 227) while workers were traversing through a secondary containment access airlock. The workers recognized the airlock condition, closed both doors within a few (less than 10) seconds and verified that the doors were latched, and notified the control room. The momentary opening of both doors within the airlock resulted in the station failing to meet TS Surveillance Requirement 3.6.4.1.2 to verify that either door in each secondary containment access opening are closed, and therefore momentarily rendered secondary containment inoperable per TS Limiting Condition for Operation 3.6.4.1. The licensee investigated the airlock doors and identified that the door 227 magnet needed adjustment. The repairs were made and the secondary containment airlock verification surveillance test was performed satisfactorily to demonstrate functionality of the interlock.

The inspectors reviewed LER 05000331/2014-003-00 against reporting requirements and found no issues. The inspectors also reviewed the licensees assessment of safety consequences in the LER, specifically the basis for not considering the condition a safety system functional failure. Based on the licensees post-LOCA dose calculation of record that did not credit secondary containment integrity for on-site and off-site doses for the first 5 minutes of the event, the inspectors determined that it was reasonable to conclude that the simultaneous opening condition of the secondary containment doors was bounded by the existing licensing basis calculation of record. Documents reviewed are listed in the Attachment to this report. This LER is closed.

This inspection constituted one event follow up sample as defined in IP 71153-05.

.4 (Closed) LER 05000331/2014-06-00: Implementation of Enforcement Guidance

Memorandum (EGM)11-003, Revision 2 Between October 12, 2014, and October 13, 2014, DAEC personnel performed operations with the potential to drain the reactor vessel (OPDRV) activities while in Mode 5 with secondary containment inoperable. An OPDRV is any activity that could result in the draining or siphoning of the RPV water level below the top of the active fuel, without taking credit for mitigating measures. Technical Specification 3.6.4.1, Secondary Containment, requires secondary containment to be operable during OPDRV activities.

Should secondary containment be found inoperable during OPDRV activities, TSs require DAEC to initiate action to suspend all OPDRV activities immediately. Therefore, performing the OPDRV activities without establishing secondary containment integrity was considered a condition prohibited by TS.

The NRC issued Enforcement Guidance Memorandum 11-003, Revision 2, on December 13, 2013, to provide guidance on how to disposition boiling water reactor licensee noncompliance with TS secondary containment requirements during OPDRV activities. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to:

(1) adhere to the NRC plain language meaning of OPDRV activities;

(2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times;

(3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a)monitoring RPV level to identify the onset of a loss of inventory event,

(b) maintaining the capability to isolate the potential leakage paths,

(c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and

(d) prohibiting movement of irradiated fuel with the spent fuel storage gates removed in Mode 5; and

(4) ensure that licensee follow all other Mode 5 TS requirements for OPDRV activities.

The inspectors reviewed this LER for potential performance deficiencies and/or violations of regulatory requirements. The inspectors also reviewed, and some instances observed, the licensees implementation of the EGM during OPDRVs:

1. The inspectors observed that the OPDRV activities were logged in the narrative logs and that the log entry appropriately recorded the standby source of makeup designated for the evolution.

2. The inspectors noted that the reactor vessel water level was maintained at least 21 feet and 1 inch above the top of the RPV flange as required by TS 3.9.7, Residual Heat Removal (RHR) - High Water Level. The inspectors also verified that at least one safety-related pump, specifically the A CS pump and associated A SBDG to provide power, was the standby source of makeup designated in the narrative logs for the evolutions. The inspectors verified the licensees calculation for worst case estimated time to drain the reactor cavity to the RPV flange was greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. The inspectors reviewed the licensees time-to-drain-down calculation during the activities and the feasibility of pre-planned actions that the station would take to isolate potential leakage paths.

4. The inspectors verified that the OPDRVs were not conducted during Mode 4 and that the licensee did not move irradiated fuel during the OPDRVs. The inspectors noted that DAEC had in place a contingency plan for isolating the potential leakage path and verified that two independent means of measuring RPV level were available, one of which had alarm indication in the control room, for identifying the onset of loss of inventory events. The inspectors further verified that secondary closure plans were in place, with required materials staged, to close up the secondary penetrations before water level reached the top of the RPV flange.

Technical Specification 3.6.4.1 required, in part, that secondary containment be operable during OPDRVs. Technical Specification 3.6.4.1, Condition C, required the licensee to initiate action to suspend OPDRVs immediately when secondary containment is inoperable. Contrary to the above, between October 12, 2014, and October 13, 2014, DAEC performed OPDRV activities while in Mode 5 without an operable secondary containment. Specifically, DAEC replaced four local power range monitors which were attached at the bottom of the reactor vessel while secondary containment was inoperable due to several secondary containment penetrations being open for scheduled cable replacements.

Because the violation occurred during the discretion period described in EGM 11-003, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and therefore, will not issue enforcement action for this violation, subject to a timely license amendment request being submitted.

In accordance with EGM 11-003, Revision 2 (ML13177A128), each licensee that receives discretion must submit a license amendment request within 12 months of the NRC staffs issuance in the Federal Register of the notice of availability for the generic change to the standard TS. The inspectors noted that DAEC is tracking the LAR submission in its CAP as CR 01998410.

Documents reviewed are listed in the Attachment to this report. This LER is closed.

This event follow-up review constituted one sample as defined in IP 71153-05.

4OA5 Other Activities

.1 Inspection of Commercial-Grade Dedication Programs

a. Inspection Scope

The inspectors reviewed policies, procedures, records, and interviewed personnel associated with the dedication of commercial grade items to verify that activities associated with dedication of those items met applicable regulatory requirements. The inspectors reviewed the licensees engineering evaluation records of commercial grade items to verify the evaluations identified the associated safety functions of the items, postulated failure modes that may adversely affect the safety functions, effects of the equipment failures, and critical characteristics of the equipment. The inspectors also assessed whether the set of critical characteristics selected for verification were sufficient to provide reasonable assurance that the item could perform its safety function.

The inspectors assessed the adequacy of the methods chosen by the licensee to verify the selected critical characteristics, including the adequacy of the identified acceptance criteria. For items that required seismic or environmental qualification, the inspectors reviewed the basis for establishing the qualification, including the basis for establishing similarity to previously tested components as applicable. Lastly, the inspectors reviewed receipt inspection and testing records associated with the selected items.

The inspectors observed testing related to commercial grade dedication associated with the following purchase orders (POs):

• 02328331 for a 24 Vdc relay; and

• 02321318 for a 125 Vdc Agastat relay.

In addition, the inspectors reviewed the dedication packages associated with the following POs:

• 02260416 for an electric control pressure switch;

• 02292658 for a capacitor;

• 02292669 for a temperature controller;

• 02293956 for electrical tape;

• 02301114 for a 300 Amp fuse;

• 02309726 for a HFD3020L circuit breaker and a fuse block;

• 02303145 for a pneumatic operator;

• 02306317 for a valve;

• 02312858 for Mobil oil;

• 02313739 for a swing check valve;

• 02320606 for an AMOT controls, thermostatic;

• 02321226 for a three way switching valve and 24V lead acid battery;

• 02321427 for a thermal overload relay;

• 02325767 for a splice;

• 02326501 for a capacitor;

• 02326544 for a solenoid valve; and

• 02328061 for a mechanical plug.

This was a pilot inspection effort and therefore, did not constitute any inspection samples.

b. Findings

Failure to Test or Evaluate the Seismic Critical Characteristic for a Commercial Grade Circuit Breaker

Introduction:

The inspectors identified a finding of very low safety significance (Green)and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control,"

for the licensee's failure to verify commercial grade circuit breakers were suitable for use in safety related applications. Specifically, the licensee failed to verify, either through seismic testing or justification, that the circuit breakers being dedicated on PO 02309726 would be able to perform their intended safety function during a seismic event.

Description:

The licensee completed a commercial grade dedication (CGD) evaluation package for PO 02309726 for commercial grade motor control center circuit breakers.

The CGD evaluation was for replacement breaker Cutler-Hammer Type HFD3020L, 20 Amps, 3-pole, 600 volt alternating current to be used as an alternate replacement for the obsolete Type HFB breaker in safety related applications.

The inspectors reviewed the purchase order and the CGD package and noticed that the licensee answered No for the seismic applicability question in the safety basis summary. In response to the inspectors questions, the licensee indicated that the No answer meant no seismic testing or seismic detailed evaluation were required for the CGD package.

The licensee indicated that the thermal-magnetic molded case circuit breakers were considered seismically rugged (not seismically sensitive) based on Generic Seismic Technical Evaluation for Replacement Items (G-STERI) Report E-95004 for molded case circuit breakers. By determination of the probabilistic seismic demand and review of critical characteristics (weight, size and location), the licensee concluded that the replacement breakers would be seismically equivalent to the breakers reviewed in the G-STERI report without evaluating if the breaker on this PO could be qualified through similarity (i.e. like-for-like or seismically equivalent) to the previously qualified breakers.

The inspectors reviewed the subsequent licensee evaluation and documentation and concluded that the licensee CGD evaluation for the replacement breaker failed to verify that the seismic characteristic for the replacement breakers was adequate. Specifically, the licensee CGD package failed to perform seismic testing for the replacement breaker and did not provide justification as to why these breakers were considered seismically rugged. The licensee used operating experience for the breaker without identifying if the breaker was like-for-like compared to the breakers tested in the G-STERI report. Since an equivalency evaluation required that the design and manufacturing techniques be considered as part of the technical justifications supporting similarity between new and previously qualified safety-related equipment, and the licensee was buying the breakers from a distributor, it was not clear how this breaker was equivalent to the breakers referenced in G-STERI Report E-95004. The inspectors determined that the licensees CGD package for the circuit breaker HFD3020L failed to provide adequate justification that would be required to demonstrate that the procured breakers were similar in the design, manufacturer material, and process as the ones previously tested or specified in the G-STERI report.

The licensee entered this issue into their CAP as CR 01986727 and performed an evaluation of a 10 percent sample of additional commercial grade dedicated molded case circuit breakers to evaluate the extent of condition. The licensee found that all the CGD evaluations sampled failed to document the justification as to why a seismic test was not required. In addition, the licensee issued CR 01987616 and concluded that these HFD3020L type circuit breakers were not yet installed at Duane Arnold and a seismic test would be performed on these type of breakers prior to installation.

Analysis:

The inspectors determined that the issue of concern represented a performance deficiency because it was the result of the licensee's failure to verify commercial grade circuit breakers were suitable for use in safety related applications.

The performance deficiency was within the licensees ability to foresee and correct and should have been prevented because the licensee had a process in place to appropriately identify and evaluate critical characteristics for commercial items being dedicated to ensure they could meet their intended safety function.

The performance deficiency was determined to be more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the failure to test or evaluate the seismic characteristics for a commercial grade circuit breaker prior installation could affect the reliability of the breaker to perform its intended safety function in a seismic event.

The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, to this finding. The inspectors answered No to all questions within Table 3, SDP Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. Per Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that because the finding did not represent an actual loss of function (circuit breakers were not currently installed), the finding screened as very low safety significance (Green).

The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of licensee's current performance.

Specifically, the licensee assumption that these types of circuit breakers were seismically rugged was based on their review of the G-STERI Technical Evaluation E-95004 which was revised in September 30, 2008.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety related functions of the structures, systems and components.

Contrary to the above, on November 1, 2013, the licensee failed to perform suitable testing or justification for a seismically sensitive component. Specifically, in a commercial grade dedication evaluation for PO 02309726 for HFD3020L, 3-poles, 20-Amp circuit breaker, DAEC personnel failed to perform seismic testing or provide a justification to show that the safety-related circuit breakers on this PO were suitable replacement parts and would perform their intended safety function during a seismic event.

This violation is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was of very low safety significance and was entered into the licensees CAP as CRs 01986727 and 01987616 to perform a seismic test prior to installation of the affected breakers. (NCV 05000331/2015001-09, Failure to Test or Evaluate the Seismic Critical Characteristic for a Commercial Grade Circuit Breaker).

4OA6 Management Meetings

.1 Exit Meeting Summary

On January 8, 2015, the inspectors presented the inspection results to Mr. T. Vehec, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

.2 Interim Exit Meetings

Interim exits were conducted for:

• The inspection results for the areas of radiological hazard assessment and exposure controls; in-plant airborne radioactivity control and mitigation; occupational dose assessment; and reactor coolant system specific activity, occupational exposure control effectiveness, and RETS/ODCM radiological effluent occurrences performance indicator verification with Mr. T. Vehec, Site Vice President, on October 31, 2014;

• The annual review of emergency action level and emergency plan changes with Mr. L. Helms, Emergency Preparedness Coordinator, via telephone on December 11, 2014;

• The results of the inservice inspection with Mr. D. Church, Engineering Programs Manager, on October 17, 2014; and

• The 2014 commercial grade dedication pilot inspections results with NextEra dedication facility and DAEC site management during on-site inspection at NextEra dedication facility in West Palm Beach, Florida, on August 29, 2014, and via telephone on December 1, 2014.

The inspectors confirmed that none of the potential report input discussed was considered proprietary. Proprietary material received during the inspection was returned to the licensee.

4OA7 Licensee-Identified Violations

The following violation of very low significance (Green) or Severity Level IV was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV.

• 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established for the selection and review for suitability of application of materials and parts that are essential to the safety-related functions of structures, systems, and components.

Contrary to the above, in October 2012, the licensee failed to properly select and review the suitability of application of several electrical cable splices and terminal strips during the replacement of safety-related electrical cables associated with the A and B SBDGs. Specifically, two modification packages associated the cable replacements during the 2012 RFO 23 did not appropriately evaluate the impacts of the stations environmental qualification (EQ) program and the effects of an internal turbine building flood.

Because the SSCs maintained operability based on the deficiency affecting the qualification of the SSCs, the finding screened as very low safety significance (Green). This issue was documented in the licensees CAP as CR 01979556.

Immediate corrective actions included a determination of operability (the SBDGs had no specified safety function for the EQ and turbine building flood events in question per the UFSAR), installation of a temporary flood barrier to compensate for the non-conforming condition for the A SBDG, cable splice and terminal strip replacement for the B SBDG, and the performance of a root cause evaluation.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. Vehec, Site Vice President

G. Pry, Plant General Manager

K. Kleinheinz, Site Engineering Director

M. Davis, Emergency Preparedness and Licensing Manager

J. Mahan, Nuclear Oversight Manager

R. Wheaton, Operations Director

R. Porter, Radiation Protection Manager

D. Olsen, Chemistry Manager

J. Schwertfeger, Security Manager

C. Hill, Training Manager

B. Murrell, Licensing Engineer Analyst

L. Swenzinski, Licensing Engineer

C. Casey, Chemistry Supervisor

L. Helms, Emergency Preparedness Coordinator

D. Church, Engineering Programs Manager

F. Dohmen, Site Level III

Nuclear Regulatory Commission

C. Phillips, Acting Chief, Reactor Projects Branch 1

M. Chawla, Project Manager, NRR

LIST OF ITEMS

Opened

05000331/2014005-01 NCV Construction Code Used during a Replacement Activity Not Reconciled with the Owners Requirements (Section 1R08.1.b.1

05000331/2014005-02 NCV Liquid Penetrant Testing Procedures Not Qualified for their Full Applicability Range (Section 1R08.1.b.2

05000331/2014005-03 NCV Failure to Evaluate Maintenance Activities for Preconditioning (Section 1R15.b.1)

05000331/2014005-04 NCV Failure to Accomplish Procedure for Leaking Pipe Snubber (Section 1R15.b.2)

05000331/2014005-05 NCV Inadequate Containment Isolation Valve Leak Tightness Test Procedure (Section 1R22.b)

05000331/2014005-06 NCV Failure to Determine Dose Rates Prior to Entry into a High Radiation Area (Section 2RS1.2.b)

05000331/2014005-07 NCV Ineffective Radiological Engineering Controls Resulted in Unplanned and Unintended Radiological Intakes to Workers (Section 2RS3.2.b)

05000331/2014005-08 NCV Failure to Evaluate Several Safety Related Relays Installed Beyond Their Design Life (Section 4OA2.3.b)

05000331/2014005-09 NCV Failure to Test or Evaluate the Seismic Critical Characteristic for a Commercial Grade Circuit Breaker (Section 4OA5.1)

Closed

05000331/2014005-01 NCV Construction Code Used during a Replacement Activity Not Reconciled With the Owners Requirements (Section 1R08.

1.b.1

05000331/2014005-02 NCV Liquid Penetrant Testing Procedures Not Qualified for their Full Applicability Range (Section 1R08.1.b.2

05000331/2014005-03 NCV Failure to Evaluate Maintenance Activities for Preconditioning (Section 1R15.b.1)

05000331/2014005-04 NCV Failure to Accomplish Procedure for Leaking Pipe Snubber (Section 1R15.b.2)

05000331/2014005-05 NCV Inadequate Containment Isolation Valve Leak Tightness Test Procedure (Section 1R22.b)

05000331/2014005-06 NCV Failure to Determine Dose Rates Prior to Entry into a High Radiation Area (Section 2RS1.2.b)

05000331/2014005-07 NCV Ineffective Radiological Engineering Controls Resulted in Unplanned and Unintended Radiological Intakes to Workers (Section 2RS3.2.b)

05000331/2014005-08 NCV Failure to Evaluate Several Safety Related Relays Installed Beyond Their Design Life (Section 4OA2.3.b)

05000331/2014005-09 NCV Failure to Test or Evaluate the Seismic Critical Characteristic for a Commercial Grade Circuit Breaker (Section 4OA5.1)

05000331/2013-006-00 LER Both Doors in Secondary Containment Airlock

Opened

Concurrently (Section 4OA3.1)

05000331/2014-002-00 LER Both Doors in Secondary Containment Airlock

Opened

Concurrently (Section 4OA3.2)

05000331/2014-003-00 LER Both Doors in Secondary Containment Airlock

Opened

Concurrently (Section 4OA3.3)

05000331/2014-006-00 LER Implementation of Enforcement Guidance Memorandum (EGM) 11-003, Revision 2 (Section 4OA3.4)

Discussed

None