IR 05000482/2015001

IR 05000482/2015001; on 01/01/2015 - 03/28/2015; Wolf Creek Generating Station Integrated Resident and Regional Report: Maintenance Risk Assessments and Emergent Work Control, and Operability Determinations and Functionality Assessments

ML15120A088
Person / Time
Site:	Wolf Creek
Issue date:	05/07/2015
From:	Andrew Rosebrook NRC/RGN-IV/DRP/RPB-B
To:	Heflin A Wolf Creek
Rosebrook A
References
IR 2015001
Download: ML15120A088 (56)

Text

UNITED STATES May 7, 2015

SUBJECT:

WOLF CREEK GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000482/2015001

Dear Mr. Heflin:

On March 28, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Wolf Creek Generating Station. On April 1, 2015, the NRC inspectors discussed the results of this inspection with you and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

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

All of these findings involved violations of NRC requirements. Further, inspectors documented one licensee-identified finding which was determined to be of very low safety significance (Green) in this report. The NRC is treating these violations as non-cited violations (NCVs)

consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Wolf Creek Generating Station.

If you disagree with a cross-cutting aspect assignment with a regulatory requirement 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 IV; and the NRC resident inspector at the Wolf Creek Generating Station.

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

Sincerely,

/RA/

Andrew A Rosebrook, Acting Chief Project Branch B Division of Reactor Projects Docket Nos. 50-482 License Nos. NPF-42

Enclosure:

Inspection Report 05000482/2015001 w/ Attachment 1: Supplemental Information Attachment 2: Request for Information for O

REGION IV==

Docket: 05000482 License: NPF-42 Report: 05000482/2015001 Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane NE Burlington, Kansas Dates: January 1 through March 28, 2015 Inspectors: C. Henderson, Acting Senior Resident Inspector D. Dodson, Acting Senior Resident Inspector R. Stroble, Resident Inspector L. Carson II, Senior Health Physicist J. Drake, Senior Reactor Inspector G. Guerra, CHP, Emergency Preparedness Inspector J. ODonnell, Health Physicist F. Thomas, Project Engineer Approved Andrew A. Rosebrook By: Acting Chief, Project Branch B Division of Reactor Projects-1- Enclosure

SUMMARY

IR 05000482/2015001; 01/01/2015 - 03/28/2015; WOLF CREEK GENERATING STATION;

Integrated Resident and Regional Report; Maintenance Risk Assessments and Emergent Work Control, and Operability Determinations and Functionality Assessments.

The inspection activities described in this report were performed between January 1 and March 28, 2015, by the resident inspectors at Wolf Creek Generating Station and inspectors from the NRCs Region IV office and other NRC offices. Three findings of very low safety significance (Green) are documented in this report. All of these findings involved violations of NRC requirements. Additionally, NRC inspectors documented in this report one licensee-identified violation of very low safety significance. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects within the Cross-Cutting Areas.

Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the failure to properly preplan maintenance such that it would not affect safety-related equipment in accordance with procedure AP 22C-008, On-Line Qualitative Risk Management, Revision 3. Specifically, during planning of emergent work activities on January 29, 2015, the licensee failed to recognize that when electrical cabinet doors containing safety-related under voltage and under frequency relays were opened to accomplish troubleshooting activities, the cabinet was not in a seismically qualified configuration. Thus the maintenance had the potential to impact the reliable operation of emergency diesel generator B during a seismic event. The licensee initiated Standing Order 37, Safety Related Cabinet Operability Requirements, Revision 0, to provide the requirements for assessing operability of opening safety-related electrical cabinet and panel doors out of their seismically qualified configuration during maintenance activities and entered this issue into their corrective action program for resolution as Condition Reports 91501 and 94605.

The licensees failure to properly preplan maintenance such that it would not affect safety-related equipment during emergent work activities was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). Specifically, the licensees failure to properly preplan maintenance resulted in emergency diesel generator B being placed in a condition that did not meet its seismic design requirements. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding:

(1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work [H.5]. (1R13)

Green.

The inspectors identified non-cited violation of 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, associated with the licensees failure to complete an adequate operability evaluation in accordance with procedure AP-28001,Opeability Evaluations, Revision 24 following the failure to meet a surveillance test acceptance criteria. Specifically, the licensee did not have an accurate technical basis for declaring the train A control room air condition unit operable when the minimum air flow rate was not met.

The licensees operability evaluation, which declared the train A control room air condition unit operable, incorrectly applied instrument uncertainty and used a superseded minimum air flow value. When these inaccuracies were addressed, the licensee determined the train was inoperable. The licensee entered this issue into their corrective action program as Condition Report 92274.

The licensees use of an inadequate technical basis for an operability evaluation of a non-conforming condition resulting in the train A control room air conditioning air condition unit being declared operable when it was actually inoperable was a performance deficiency.

The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associate cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with conservative bias component because the licensee did not use a decision making-practice that emphasized prudent choices over those that are simply allowable. A proposed action was determined to be safe in order to proceed, rather than unsafe in order to stop [H.14]. (1R15.1)

Cornerstone: Barrier Integrity

Green.

The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, Drawings, associated with the licensees failure to follow the requirements of Station Procedure AP 10-104, Breach Authorization, Revision 32.

Specifically, the licensees failure initiate a breach permit and station a boundary watch when the auxiliary building emergency exhaust system boundary door 41015 was opened multiple times for transporting scaffolding from the turbine building to the auxiliary building.

Opening this door without compensatory measures rendered the auxiliary building emergency exhaust system inoperable. The license entered this issue into their corrective action program for resolution as Condition Reports 92315 and 92630.

The licensees failure to initiate a breach permit and implement required compensatory measures for when the auxiliary building emergency exhaust system boundary door 41015 was open was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the system, structure, and component and barrier performance attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the radiological barrier functionality of the auxiliary building emergency exhaust system. Specifically, without a dedicated individual in constant communication with the control room, as required by AP 10-104, opening this door required entry of Technical Specification 3.7.13 Limited Condition of Operation Condition B.

The longest period door 41015 was open was approximately one hour without the required compensatory measure. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding screened as having very low safety significance (Green) because the finding only involved a degradation of the radiological barrier function provided for the auxiliary building. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work [H.5]. (1R15.2)

Licensee-Identified Violations

A violation of very low safety significance (Green) was identified by the licensee and was reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and associated corrective action tracking numbers are listed in Section 4OA7 of this report.

PLANT STATUS

Wolf Creek began the inspection period at 100 percent power. On February 28, 2015, the unit was shut down for Refueling Outage 20. The unit remained shutdown for the remainder of the inspection period.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On January 8, 2015, the inspectors completed an inspection of the stations readiness for impending adverse weather conditions. The inspectors reviewed plant design features, the licensees procedures to respond to cold weather on the circulation water screen house cold weather compensatory measures, and the licensees implementation of these procedures. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant.

These activities constituted one sample of readiness for impending adverse weather conditions, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

1R04 Equipment Alignment

Partial Walkdown

a. Inspection Scope

The inspectors performed partial system walk-downs of the following risk-significant systems:

January 6, 2015, turbine-driven auxiliary feedwater January 21, 2015, containment spray B January 26, 2015, essential service water A January 27, 2015, centrifugal charging pump A The inspectors reviewed the licensees procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the systems were correctly aligned for the existing plant configuration.

These activities constituted four partial system walk-down samples as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

1R05 Fire Protection

Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensees fire protection program for operational status and material condition. The inspectors focused their inspection on four plant areas important to safety:

February 20, 2015, service water February 28, 2015, reactor building fire area RB, containment March 10, 2015, essential service water fire area A, essential service water pump house, Room A March 10, 2015, fuel building fire area F-1, spent fuel pool For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensees fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions.

These activities constituted four quarterly inspection samples, as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

a. Inspection Scope

On March 11, 2015, the inspectors completed an inspection of the stations ability to mitigate flooding due to internal causes. After reviewing the licensees flooding analysis, the inspectors chose plant areas containing risk-significant structures, systems, and components that were susceptible to flooding:

Engineered safety features Class 1E switchgear NB01/NB02 rooms 3301 and 3302 and Operating Experience Smart Sample 2007-02, Flooding Vulnerabilities Due to Inadequate Design and Conduit/Hydrostatic Seal Barrier Concerns The inspectors reviewed plant design features and licensee procedures for coping with internal flooding. The inspectors walked down the selected areas to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether operator actions credited for flood mitigation could be successfully accomplished.

In addition, on March 16, 2015, the inspectors completed an inspection of underground bunkers susceptible to flooding. The inspectors selected three vaults that contained risk-significant or multiple-train cables whose failure could disable risk-significant equipment:

Essential service water A vault 2A Essential service water A vault 3A Essential service water A vault 4A The inspectors observed the material condition of the cables and splices contained in the vaults and looked for evidence of cable degradation due to water intrusion. The inspectors verified that the cables and vaults met design requirements.

These activities constitute completion of one flood protection measures sample and one bunker/manhole sample, as defined in Inspection Procedure 71111.06.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

The activities described in subsections 1 through 5 below constitute completion of one inservice inspection sample, as defined in Inspection Procedure 71111.08.

.1 Non-destructive Examination (NDE) Activities and Welding Activities

a. Inspection Scope

The inspectors directly observed the following nondestructive examinations:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Reactor Coolant TBB03-10C-IR Ultrasonic System Reactor Coolant TBB03-10C-W Ultrasonic System Reactor Pressure Ultrasonic RV-LIG-12 Vessel Ligaments Chemical and PW1A Penetrant Volume Control System SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Chemical and PW1A Visual Volume Control System Reactor Pressure RBB01Flange Mating Surface Visual Vessel Reactor Pressure RBB01Upper Support Plate Mating Visual Vessel Surface Reactor Pressure Hot Leg Nozzle to Pipe Visual Vessel The inspectors reviewed records for the following nondestructive examinations:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Reactor Coolant TBB03-10C-IR Ultrasonic System Reactor Coolant TBB03-10C-W Ultrasonic System Reactor Pressure RV-LIG-12,37-54 Ultrasonic Vessel Ligaments Accumulator Safety EP-01-S003-K Ultrasonic Injection System Accumulator Safety EP-01-S003-L Ultrasonic Injection System Chemical and BG-02-H007 Visual Volume Control System During the review and observation of each examination, the inspectors observed whether activities were performed in accordance with the ASME Code requirements and applicable procedures. The inspectors also reviewed the qualifications of all nondestructive examination technicians performing the inspections to determine whether they were current.

The inspectors directly observed a portion of the following welding activities:

SYSTEM WELD IDENTIFICATION WELD TYPE Chemical and Volume PW-1A Gas Tungsten Arc Control System Welding SYSTEM WELD IDENTIFICATION WELD TYPE Chemical and Volume SW-2A Gas Tungsten Arc Control System Welding Essential Service FW-500 Shielded Metal Arc Water Welding High Pressure FW-1 Gas Tungsten Arc Coolant Injection Welding System The inspectors reviewed records for the following welding activities:

SYSTEM WELD IDENTIFICATION WELD TYPE Fuel Pool Cooling & PW-1A Gas Tungsten Arc Clean-Up System Welding Fuel Pool Cooling & PW-2 Gas Tungsten Arc Clean-Up System Welding Fuel Pool Cooling & PW-3 Gas Tungsten Arc Clean-Up System Welding Essential Service W-3 Gas Tungsten Arc Water Welding Essential Service W-4 Gas Tungsten Arc Water Welding Essential Service W-6 Gas Tungsten Arc Water Welding Essential Service W-8A Gas Tungsten Arc Water Welding Essential Service W-9A Gas Tungsten Arc Water Welding The inspectors reviewed whether the welding procedure specifications and the welders had been properly qualified in accordance with ASME Code Section IX requirements.

The inspectors also determined that essential variables were identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specifications.

b. Findings

No findings were identified, but the following item is unresolved.

Introduction.

The inspector identified an unresolved item pertaining to 10 CFR 50 Appendix B, Criterion IX, Control of Special Processes, associated with the licensees method of performing ultrasonic examination of the reactor vessel flange stud hole threads in accordance with applicable American Society of Mechanical Engineers (ASME) Code requirements.

Description.

The inspector identified several issues of concern while observing the licensees ultrasonic examination of the reactor vessel flange stud hole threads. The inspector questioned whether the licensee would be able to detect any reportable indication within the ASME Code examination zone using the technique employed.

The inspector identified that in 2003 the licensee had modified the method used to perform the examination scanning, but never verified that the new methodology was capable of detecting relevant indications within the examination zone.

The new method placed the one inch diameter zero angle transducer on a radial arm at the end of an approximately 30 foot pole. The pole is aligned on the handle of the protective cap that covers the stud hole in the flange.

The inspector reviewed examination Procedure UT-11, Ultrasonic Stud Hole Threads, Revision 13, and Examination of Reactor Vessel Flange made note of the following:

The inner edge of the transducer is at a nominal distance of 3.875 inches from the center of the stud hole.

The protective cap has a nominal diameter of 7.25 inches or a radius of 3.625 inches while the stud hole diameter is 6.822 +0, -.01 inches.

This places the examination zone of inspection starting at a radius of approximately 3.411 inches and extending to a radius of 4.411 inches.

The configuration of the transducer on the pole and the alignment mechanism results in the inside edge of the transducer being placed approximately 0.465 inches from the edge of the stud hole, which is the start of the one inch examination area. Because the technique employs a "zero" angle transducer and the examination area is not directly beneath the transducer, there is a concern with instrument signal coverage.

The inspector also identified several procedural compliance issues while reviewing the licensees implementation of UT-11. The inspector questioned the following statements in the procedure:

Procedure UT-11, Section 11.1.1, states in part, The examination volume is a one inch annular band around each stud hole, extending to one stud diameter into the flange.

Procedure UT-11, Section 11.2.2, states in part that, Straight beam examination of ligaments shall be performed.

Procedure UT-11, Section 12.1.1, states in part, All indications which are found that are orientated on a plane normal to the axis of the stud that are equal to or exceed 0.2 in, as measured radially from the root of the thread, shall be reported to the LMT Site Supervisor and recorded on the Ultrasonic Examination report form.

There is a concern that the technique currently being utilized by the licensee may not provide adequate coverage of the required examination area and may not be capable of detecting indications orientated on a plane normal to the axis of the stud that are equal to or exceed 0.2 inch, as measured radially from the root of the thread, as required by the licensee's procedure and Section XI of the ASME Code. Additional analysis and simulations need to be completed to determine if the licensee is meeting ASME Code requirements. This issue is being tracked as URI 05000482/2015001-01, Questions Related to Ultrasonic Examination of Reactor Vessel Flange Stud Hole Threads.

.2 Vessel Upper Head Penetration Inspection Activities

a. Inspection Scope

The inspectors reviewed the results of the licensees bare metal visual inspection of the reactor vessel upper head penetrations to determine whether the licensee identified any evidence of boric acid challenging the structural integrity of the reactor head components and attachments. The inspectors also verified that the required inspection coverage was achieved and limitations were properly recorded.

During refueling outage RF19, ultrasonic examinations of all seventy-eight control rod drive mechanism (CRDM) penetration nozzles and the eddy current examination of the vent line in the reactor vessel head was completed. A number of thermal sleeves were found to have wear indications extending up to as much as 360 degrees around the thermal sleeve where the thermal sleeve exits the bottom end of the control rod drive mechanism head adapter tube. Wear was found in rodded and unrodded penetration locations. The wear is attributed to the thermal sleeve contacting the inside diameter of the CRDM head adapter tube due to a flow-induced impact/whirling motion of the thermal sleeve. The sleeve-to-adapter contact resulted in wear of material on the outside diameter of the thermal sleeves. A sample of the thermal sleeves were re-inspected this outage and no change in the wear indications were noted.

During refueling outage RF20, a visual examination (VT-2) of the reactor pressure vessel head was performed. The examination was in accordance with Code Case N 729-1 Table 1, Item B4.20. An indication of primary water stress corrosion cracking was identified on the canopy seal weld for CRDM penetration 20. The CRDMs were fabricated in sections with threaded joints providing the pressure-retaining capabilities. Since the threaded joint provides pressure retention, the canopy seal weld is not pressure retaining and is for leakage control. The licensee installed a mechanical clamp on the canopy seal weld to restore leakage control.

The inspectors reviewed the certification of the personnel performing the inspection to verify they were certified examiners to their respective nondestructive examination method.

b. Findings

No findings were identified.

.3 Boric Acid Corrosion Control (BACC) Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensees implementation of its boric acid corrosion control program for monitoring degradation of those systems that could be adversely affected by boric acid corrosion. The inspectors reviewed the documentation associated with the licensees boric acid corrosion control walk-down as specified in Procedures STN PE-040D and AI 16F-002. The inspectors also reviewed the visual records of the components and equipment. The inspectors verified that the visual inspections emphasized locations where boric acid leaks could cause degradation of safety-significant components. The inspectors also verified that the engineering evaluations for those components where boric acid was identified gave assurance that the ASME Code wall thickness limits were properly maintained. The inspectors confirmed that the corrective actions performed for evidence of boric acid leaks were consistent with requirements of the ASME Code and 10 CFR 50, Appendix B requirements. Specific documents reviewed during this inspection are listed in the attachment.

b. Findings

No findings were identified.

.4 Steam Generator Tube Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensees in-situ pressure testing screening criteria for flawed steam generator tubes to verify that it was in accordance with the EPRI guidelines. The inspectors also reviewed the steam generator tube eddy current examination scope and expansion criteria to verify that these meet technical specification requirements. The inspector reviewed the licensees inspection of the secondary side of the steam generators, and corrective actions taken in response to any observed degradation. The licensee did repairs on select tubes (e.g., installed plugs),and the inspectors observed a portion of these repairs. The inspector observed the licensees vendor to determine if the equipment was qualified for detection and/or sizing of the expected types of tube degradation. The inspectors observed the licensees vendor performing analysis of the steam generator tubes to determine if proper eddy current testing analysis techniques were applied.

The primary side inspection scope performed in all four steam generators for the current outage, RF20, included the following:

25 percent Bobbin examination of tubes in all four steam generators 25 percent hot leg rotating pancake coil (RPC) Tube Sheet (TS) +3"/-15.21" Cold Leg Peripheral Tubes, Tube Sheet Cold (TSC) +/- 3" 100 percent of peripheral tubes

+Point examination of all "1-code" indications not resolved after history review

+Point inspection to bound the tubes with possible loose part signals

+Point inspection of possible loose part signals from the previous inspection as specified in Section 3.5 25 percent Row 1 and Row 2 U-bends, mid-range +Point examination Dents (structures) >5 volts: Inspect 50 percent in steam generator Band C, and 25 percent in steam generators A and D of all previously identified and all new dents >5 volts in the hot leg (including the U-bends) with the mid-range +Point probe in all four SGs Dings (free span) >5 volts: Inspect 25 percent of all previously identified and all new dings >5 volts in the hot leg (including the U-bends) with the mid-range

+Point probe in all four Steam Generators. A "new" ding is defined as one for which there is no prior historical record 100 percent Bobbin inspection of all prior indications except dents and dings

+Point examination of a 5 percent sample of bobbin indications that have not changed since the prior inspection ("H" and "S" codes)

+Point inspection of the sample of tubes to support the scale profiling effort I00 percent bobbin inspection of tubes identified as potentially having high residual stress 100 percent bobbin inspection of active tubes surrounding previously plugged tubes Visual inspections of all plugs, including factory installed plugs, or their replacements Inspection of potentially deleterious foreign objects During the initial eddy current examinations in steam generator A, a single circumferential indication was identified in the hot leg tube sheet of tube (R20, C102)approximately 4 inches down from the top of the tube sheet. This indication is not located within a specified examination subset of the hot leg tube sheet (bulge or overexpansion). This primary water stress corrosion cracking (PWSCC) indication is associated with a low level (4.0 volt) bulge anomaly that is below the threshold of the bulge signal reporting criteria (18 volts) that had not previously been identified as a degradation mechanism. The tube was plugged and because the indication is 4 inches inside the tube sheet, there are no concerns with lateral movement resulting in tube severance if the indication grows. Because the tube is unpressurized, there is no pull-out force to cause vertical motion. Therefore, there was no need to stabilize the tube.

The current EPRI Steam Generator Examination Guidelines for this damage mechanism require that a 100 percent inspection of affected steam generator (steam generator A)and a 20 percent inspection in the unaffected steam generators (steam generators B, C, and D) be completed. Wolf Creek expanded the hot leg top of tube sheet eddy current examinations to 100 percent in steam generator A and a minimum of 50 percent in steam generators B, C, and D. No additional indications were identified in the expanded scope inspection.

The inspectors reviewed the licensees known tube degradation mechanisms.

The inspectors observed portions of the eddy current testing being performed to determine whether:

(1) the appropriate probes were used for identifying the expected types of degradation,

(2) calibration requirements were adhered, and

(3) probe travel speed was in accordance with procedural requirements. The inspectors performed a review of the site-specific qualifications for the techniques being used and reviewed whether eddy current test data analyses were adequately performed per EPRI and site specific guidelines. The inspectors selected a number of degraded tubes and compared them to the previous outage operational assessment to assess the licensees prediction capabilities. As a result of the eddy current inspection, thirty-one tubes were plugged during RF20.

Finally, the inspectors reviewed selected eddy current test data to verify that the analytical techniques used were adequate.

The inspectors reviewed the licensees actions in response to six metallic objects identified in the steam generators. The licensee was able to retrieve three of the objects and the remaining objects were evaluated as satisfactory to remain in place

b. Findings

No findings were identified.

.5 Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed of a sample of problems associated with inservice inspections documented by the licensee in the corrective action program for appropriateness of the corrective actions. 15 condition reports were selected which dealt with inservice inspection activities and found the corrective actions were appropriate. The specific condition reports reviewed are listed in the documents reviewed section. From this review, the inspectors concluded that the licensee has an appropriate threshold for entering issues into the corrective action program and has procedures that direct a root cause evaluation when necessary. The licensee also has an effective program for applying industry operating experience. Specific documents reviewed during this inspection are listed in the attachment.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On January 26, 2015, the inspectors observed an evaluated simulator scenario performed by an operating crew. The inspectors assessed the performance of the operators and the evaluators critique of their performance. The inspectors also assessed the modeling and performance of the simulator during the evaluated scenario.

These activities constitute completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

.2 Review of Licensed Operator Performance

a. Inspection Scope

The inspectors observed the performance of on-shift licensed operators in the plants main control room. At the time of the observations, the plant was in a period of heightened activity. The inspectors observed the operators performance of the following activities:

January 22, 2015, control rod adjustments, response to instrument tunnel sump running, including the pre-job brief February 28, 2015, reactor plant shutdown for Refueling Outage 20 In addition, the inspectors assessed the operators adherence to plant procedures, including conduct of operations procedure and other operations department policies.

These activities constitute completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed one instance of degraded performance or condition of safety-related structures, systems, and components (SSCs):

March 11, 2015, degraded floor drains and watertight doors affecting the internal flooding analysis for the auxiliary building The inspectors reviewed the extent of condition of possible common cause SSC failures and evaluated the adequacy of the licensees corrective actions. The inspectors reviewed the licensees work practices to evaluate whether these may have played a role in the degradation of the SSCs. The inspectors assessed the licensees characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.

These activities constituted completion of one maintenance effectiveness sample, as defined in Inspection Procedure 71111.12.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed five risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk:

January 14, 2015, planned maintenance on the motor-driven and diesel-driven fire pumps with circulating water pump C out-of-service February 3, 2015, station blackout diesel generator A, emergency diesel generator B, and essential service water B out-of-service February 24, 2015, high energy line break door 41015 breached for scaffolding installation March 4, 2015, orange risk window for reactor coolant system lowered inventory March 11, 2015, spent fuel pool cooling risk management plan for train B unavailable and associated yellow risk window.

The inspectors verified that these risk assessments were performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensees risk assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments.

Additionally, on January 29, 2015, the inspectors also observed a portion of an emergent work activity for emergency diesel generator B exciter potential transformer fuse alarms that had the potential to affect the functional capability of mitigating systems.

The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on unaffected structures, systems, and components.

These activities constitute completion of six maintenance risk assessments and emergent work control inspection samples, as defined in Inspection Procedure 71111.13.

b. Findings

Failure to Assess the Operability of Emergency Diesel Generator B during Emergent Work Activities.

Introduction.

The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the licensees failure to properly preplan maintenance such that it would not adversely affect safety-related equipment in accordance with procedure AP 22C-008, On-Line Qualitative Risk Management, Revision 3. Specifically, during emergent work activities, the licensee failed to recognize that when electrical cabinet doors containing safety-related under voltage and under frequency relays were opened to accomplish maintenance, the cabinet was no longer in a seismically qualified configuration.

Description.

On January 29, 2015, while troubleshooting an intermittent power potential transformer fuse blown alarm for the emergency diesel generator B, maintenance personnel opened the doors to panel NE 106 to gain access to the relay NE 106160 per Work Order 15-397359-000. During the maintenance, inspectors noted that the doors were not restrained and there was not a dedicated person attending the door. The door associated with panel NE 106 contained safety-related under voltage and under frequency relays.

The inspectors asked the licensee if the safety-related relays where seismically qualified with the door open. The licensee informed the inspectors that the safety-related relays where not seismically qualified with the panel door open. The inspectors were concerned that in the event of a seismic event, the doors could suddenly shut and cause the relays to change state, impacting the reliability of emergency diesel generator B at a time when it was required to perform its safety function. Thus, the inspectors concluded that the licensee should have declared the emergency diesel generator inoperable and entered the appropriate technical specification limiting condition for operation prior to the commencement of the maintenance. The emergency diesel generator was not in a non-conforming configuration for greater than the technical specification allowed outage time.

The inspector reviewed Station Procedure AP 22C-008, On-Line Qualitative Risk Management, Revision 3, and determined the licensee failed to identify the worst case consequences (i.e., seismic event) and have appropriate mitigating actions for the emergent work activity in accordance with step 6.2.3 of the procedure when planning the emergent work activities for emergency diesel generator B. The licensee initiated Condition Reports 91501 and 94605 to document this issue in the corrective action program. Condition Report 91501 was initiated on February 3, 2015, for an industry concern regarding the opening of doors of operable safety related electrical cabinets and panels and Condition Report 94605 was initiated for the inspectors issue identified on January 29, 2015.

In response to Condition Report 91501 the licensee initiated Standing Order 37, Safety Related Cabinet Operability Requirements, Revision 0. The standing order outlined expectations for opening safety related electrical cabinets. Specifically the standing order required;

(1) control room permission prior to opening any safety related cabinets;

(2) the doors shall be attended at all times;

(3) the doors shall be restrained, and the doors to be shut immediately if a seismic event were to occur.

Analysis.

The failure to properly preplan maintenance such that it would not affect safety-related equipment was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associate cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage).

Specifically, the licensees failure to preplan maintenance resulted in emergency diesel generator B being placed into a condition that did not meet its seismic design requirements. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding:

(1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality;

(2) did not represent a loss of system and/or function;

(3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and

(4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work [H.5].

Enforcement.

Technical Specification 5.4.1.a requires, in part, that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, February 1978. Section 9.a of Regulatory Guide 1.33 requires maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Work Order 15-397359-000 provided work instructions for troubleshooting inside an electrical cabinet associated with emergency diesel generator B. Contrary to the above, on January 29, 2015, the licensee performed maintenance that affected safety-related equipment was not performed in accordance with documented instructions that were appropriate to the circumstances. Specifically, troubleshooting activities performed under Work Order 15-397359-000 caused the safety related emergency diesel B to be rendered non-conforming to its seismic requirements. Because the finding was of very low safety significance (Green) and was entered into the licensees corrective action program as Condition Report 94605, it is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000482/2015001-02, Failure to Assess the Operability of Emergency Diesel Generator B during Emergent Work Activities.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors reviewed four operability determinations that the licensee performed for degraded or nonconforming structures, systems, or components (SSCs):

February 13, 2015, Condition Reports 88665 and 91799, operability determination of emergency diesel generator B static exciter voltage regulator power rectifier bank diode stacking faults February 23, 2015, Condition Report 92100, operability determination of essential service water through-wall leak.

February 26, 2015, Condition Report 92109, operability determination of the train A control room air conditioning unit March 4, 2015, Condition Report 92315, operability determination of auxiliary building emergency exhaust system with door 41015 open greater than three-quarters of an inch.

The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded SSC to be operable, the inspectors verified that the licensees compensatory measures were appropriate to provide reasonable assurance of operability. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability of the degraded SSC.

These activities constitute completion of four operability and functionality review samples as defined in Inspection Procedure 71111.15.

1. Failure to Complete an Adequate Operability Evaluation for Declaring the Train A

Control Room Air Conditioning Unit Operable

Introduction.

The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to complete an adequate operability evaluation in accordance with procedure AP-28001,Opeability Evaluations, Revision 24, following the failure to meet a surveillance test acceptance criteria. Specifically, the licensee did not have an accurate technical basis for declaring the train A control room air condition unit operable when the minimum air flow rate was not met. The licensees operability evaluation, which declared the train A control room air condition unit operable, incorrectly applied instrument uncertainty and used a superseded minimum air flow value. When these inaccuracies were addressed, the licensee determined the train was inoperable.

Description.

The inspectors reviewed Condition Report 92109 based on its risk significance for maintaining of the control rooms habitability during a design basis accident. This condition report documented that the train A control room air condition unit did not meet the minimum air flow acceptance criteria in accordance with Station Procedure STS PE-010A, Control Room A/C System Flow Rate Verification A Train, Revision 3A, and included an operability determination which the licensee completed on February 19, 2015.

The inspectors reviewed the operability determination, Station Procedure STS PE-010A, and Design Calculation GK-M-001, Cooling and Heating Load Calculation for Control Room HVAC System Capabilities during Normal Plant Operation and Accident Conditions -(SGK04A/B), Revision 3.

During a surveillance test on February 25, 2015, the air flow rate measured was 20,760 cfm, which was less that the minimum air flow rate acceptance criterion of 21,012 cfm stated in the surveillance procedure. Step 4.2 of STS PE-010A required technicians to notify operations if the control room air conditioner flow rate is less than 21,012 cfm, and to refer to Limited Condition of Operation 3.7.11 for applicable action.

The inspectors determined that, rather than declaring the train inoperable and taking the actions required by technical specifications, the licensee had performed an operability evaluation that incorrectly applied instrument uncertainty and used a minimum flow rate value that had been superseded. Specifically, the inspectors determined that the surveillance acceptance criteria already accounted for instrument uncertainty, so the operability determination incorrectly applied the instrument uncertainty factor twice.

Additionally the operability evaluation used a minimum air flow value of the air condition units that was taken from Revision 2 of Calculation GK-M-001. Revision 2 was no longer the current version of the calculation and the minimum air flow value had been revised to a higher value in Revision 3 of Calculation GK-M-001.

When the inspectors brought this to the attention of the licensee, Condition Report 92274 was written to document this issue in the licensees corrective action program.

The licensee subsequently concluded that control room air condition system train A was inoperable on February 25, 2015. The flow rate was corrected by adjusting flow dampers and re-performing the test, and returned to operable status on March 6, 2015.

Analysis.

The licensees use of an inadequate technical basis for an operability evaluation of a non-conforming condition resulting in the train A control room air conditioning air condition unit being declared operable when it was actually inoperable was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associate cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding:

(1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality;

(2) did not represent a loss of system and/or function;

(3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and

(4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with conservative bias component because the licensee did not use a decision making practice that emphasized prudent choices over those that are simply allowable in that they did not determined the proposed action to be safe in order to proceed [H.14].

Enforcement.

10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality 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, and drawings.

Station Procedure AP-28001,Opeability Evaluations, Revision 24 required non-conforming conditions to be evaluated for operability. Station Procedure STS PE-010A, Control Room A/C System Flow Rate Verification A Train, Revision 3A, Step 4.2, states, notify operations that if it is determined that the control room air conditioner flow rate is less than 21,012 cfm then refer to Limited Condition of Operation 3.7.11 for applicable action for the limiting condition of operation. Specifically, the licensee completed an inadequate operability evaluation due to using incorrect data and assumptions which resulted in an inoperable system being declared operable. Because the finding was of very low safety significance (Green) and was entered into the licensees corrective action program as Condition Report 92274, it is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000482/2015001-03, Failure to Complete an Adequate Operability Evaluation for Declaring the Train A Control Room Air Conditioning Unit Operable.

2. Failure to Station Boundary Watch for Opening Auxiliary Building Emergency Exhaust

System Boundary Door

Introduction.

The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, Drawings, associated with the licensees failure follow the requirements of Station Procedure AP 10-104, Breach Authorization, Revision 32. Specifically, the licensees failure initiate a breach permit and station a boundary watch when the auxiliary building emergency exhaust system boundary door 41015 was open greater than three-quarters of an inch for other than entry and exit through the door for transporting scaffolding from the turbine building to the auxiliary building.

Description.

On February 24, 2015, the inspectors identified door 41015 was opened multiple times during plant status walk down of the auxiliary building and turbine building while scaffolding material was moved from the turbine building to the auxiliary building.

The inspector reviewed the requirements of Station Procedure AP 10-104, Breach Authorization, Revision 32, and identified that door 41015 was a fire boundary and pressure boundary for the auxiliary building emergency exhaust system. This procedure required that if a single door opening in the auxiliary building emergency exhaust system barrier envelope was planned to be open more than 3/4-inch, it would require obtaining a breach permit. Where auxiliary building emergency exhaust system barrier envelope integrity was affected, compensatory measures were required, including stationing a dedicated individual to act as a Boundary Watch to maintain barrier operability/functionality. The plant was operating in Mode 1 at the time, so the auxiliary building emergency exhaust system was required to be operable. Thus, with the door open and no compensatory measure taken, an entry into the technical specification limiting condition for operation should have been made. The door was not breeched for greater than the technical specification allowed outage time.

The inspectors informed the licensee of the issue with door 41015 and asked if a breach authorization permit was issued on February 24, 2015. The licensee determined that a breach authorization permit was not issued and initiated Condition Report 92315 into their corrective action program.

The inspectors reviewed the events leading up to the door being opened, and found that maintenance and security personnel had requested a breach permit, but operations and fire protection personnel had incorrectly concluded that a breach permit was not needed.

The inspectors determined that the licensee had not addressed all the boundary functions of door 41015 (specifically the pressure boundary function for auxiliary building emergency exhaust system), and had incorrectly applied the requirements of procedure AP 10-104, Section 6.7.1. The licensee entered this issue into their corrective action program for resolution as Condition Report CR 92630.

The inspectors determined that the performance deficiency did not impair the high energy line break or fire protection functions.

Analysis.

The failure to initiate a breach permit and take required compensatory measures prior to opening auxiliary building emergency exhaust system boundary door 41015 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the system, structure, and component and barrier performance attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the radiological barrier function of the auxiliary building emergency exhaust system. Specifically, without a dedicated individual in constant communication with the control room, as required by AP 10-104, opening this door rendered the emergency exhaust system inoperable. The longest period door 41015 was open was approximately one hour without the required compensatory measure. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding screened as having very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the auxiliary building. The finding has a cross-cutting aspect in the area of human performance associated with work management. The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities [H.5].

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality 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, and drawings. Station Procedure AP 10-104, Breach Authorization, Revision 32, section 6.3.3.2.c, Where Auxiliary Building Emergency Exhaust System barrier envelope integrity is affected, one of the following compensatory measures shall be applied:

(1) Utilize a Boundary Watch to maintain barrier operability/functionality in accordance with Section 6.11 of AP 26C-004, Operability Determination and Functionality Assessment, Revision 30. Contrary to above, on February 24, 2015, the licensee failed to follow Station Procedure AP 10-104 while breaching the auxiliary building emergency exhaust system boundary, and activity affecting quality. Specifically, the licensee did not station a boundary watch in continuous contact with the control room to be able to rapidly close the door when the auxiliary building emergency exhaust system boundary door 41015 was open. Because the violation was of very low safety significance (Green) and was entered into the licensees corrective action program as Condition Reports 92315 and 92630, it is being treated as a non-cited violation, in accordance with Section 2.3.2.a of the NRCs Enforcement Policy: NCV 05000482/2015001-04, Failure to Station Boundary Watch for Opening Auxiliary Building Emergency Exhaust System Boundary Door.

1R18 Plant Modifications

Permanent Modifications

a. Inspection Scope

On March 18, 2015, the inspectors reviewed a permanent modification to allow the ability to isolate the refueling water storage tank from the fuel pool cleanup system.

The inspectors reviewed the design and implementation of the modification. The inspectors verified that work activities involved in implementing the modification did not adversely impact operator actions that may be required in response to an emergency or other unplanned event. The inspectors verified that post-modification testing was adequate to establish the operability of the SSC as modified.

These activities constitute completion of one sample of permanent modifications, as defined in Inspection Procedure 71111.18.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed five post-maintenance testing activities that affected risk-significant structures, systems, or components (SSCs):

January 6, 2015, motor-driven auxiliary feedwater pump following planned maintenance January 27, 2015, centrifugal charging pump A following planned maintenance January 28, 2015, valve EMHV-8821A safety injection pump discharge valve following planned maintenance March 24, 2015, emergency diesel generator A over speed testing following governor shaft replacement and preventative maintenance March 27, 2015, essential service water B valves EFHV0039 and EFHV0041 service water isolation valves leak test following valve maintenance The inspectors reviewed licensing- and design-basis documents for the SSCs and the maintenance and post-maintenance test procedures. The inspectors observed the performance of the post-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected SSCs.

These activities constitute completion of five post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

During the stations refueling outage that commenced on February 28, 2015, the inspectors evaluated the licensees outage activities. The inspectors verified that the licensee considered risk in developing and implementing the outage plan, appropriately managed personnel fatigue, and developed mitigation strategies for losses of key safety functions. This verification included the following:

Review of the licensees outage plan prior to the outage Monitoring of shut-down and cool-down activities Verification that the licensee maintained defense-in-depth during outage activities Observation and review of reduced-inventory activities and mid-loop activities Observation and review of fuel handling activities These activities constitute completion of one refueling outage sample, as defined in Inspection Procedure 71111.20.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed six risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the structures, systems, and components (SSCs) were capable of performing their safety functions:

In-service tests:

March 5, 2015, emergency core cooling system check valve testing Containment isolation valve surveillance tests:

February 6, 2015, residual heat removal B to safety injection pump test line isolation valve stroke time testing, train B residual heat removal system inservice valve test February 6, 2015, safety injection test line system inside containment isolation valve, boron injection upstream test line isolation, and accumulator tank fill line isolation valve stroke-time testing, safety injection system train B inservice valve test Reactor coolant system leak detection tests:

February 13, 2015, reactor coolant system unidentified leakage calculation Other surveillance tests:

February 23, 2015, emergency diesel generator A monthly operability run March 17, 2015, control room air conditioning system A flow rate verification test The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected SSCs following testing.

These activities constitute completion of six surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.

b. Findings

No findings were identified. A finding associated with the train A control room air conditioning system flow rate verification test is documented in Section 1R15.1.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspector performed an in-office review of Wolf Creek Generating Station Emergency Plan Procedure EPP 06-007 Emergency Notifications, Revision 22. The change added instructions to make notifications to the NRC from the Technical Support Center in the event the Control Room is not habitable or has been evacuated, included guidance for obtaining meteorological data from alternate sources, and instructions on how security sensitive information should be provided to the NRC.

This revision was compared to its previous revision, to the criteria of NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, and to the standards in 10 CFR 50.47(b) to determine if the revision adequately implemented the requirements of 10 CFR 50.54(q)(3) and 50.54(q)(4). The inspector verified that the revision did not decrease the effectiveness of the emergency plan. This 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.

These activities constitute completion of one emergency action level and emergency plan changes sample as defined in Inspection Procedure 71114.04.

b. Findings

No findings were identified.

1EP6 Drill Evaluation

Training Evolution Observation

a. Inspection Scope

On January 26, 2015, the inspectors observed simulator-based licensed operator requalification training that included implementation of the licensees emergency plan.

The inspectors verified that the licensees emergency classifications, off-site notifications, and protective action recommendations were appropriate and timely. The inspectors verified that any emergency preparedness weaknesses were appropriately identified by the evaluators and entered into the corrective action program for resolution.

These activities constitute completion of one training observation sample, as defined in Inspection Procedure 71114.06.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Public Radiation Safety and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

The inspectors assessed the licensees performance in assessing the radiological hazards in the workplace associated with licensed activities. The inspectors assessed the licensees implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures. The inspectors walked down various portions of the plant and performed independent radiation dose rate measurements. The inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors reviewed licensee performance in the following areas:

The hazard assessment program, including a review of the licensees evaluations of changes in plant operations and radiological surveys to detect dose rates, airborne radioactivity, and surface contamination levels Instructions and notices to workers, including labeling or marking containers of radioactive material, radiation work permits, actions for electronic dosimeter alarms, and changes to radiological conditions Programs and processes for control of sealed sources and release of potentially contaminated material from the radiologically controlled area, including survey performance, instrument sensitivity, release criteria, procedural guidance, and sealed source accountability Radiological hazards control and work coverage, including the adequacy of surveys, radiation protection job coverage and contamination controls, the use of electronic dosimeters in high noise areas, dosimetry placement, airborne radioactivity monitoring, controls for highly activated or contaminated materials (non-fuel) stored within spent fuel and other storage pools, and posting and physical controls for high radiation areas and very high radiation areas Radiation worker and radiation protection technician performance with respect to radiation protection work requirements Audits, self-assessments, and corrective action documents related to radiological hazard assessment and exposure controls since the last inspection These activities constitute completion of one sample of radiological hazard assessment and exposure controls as defined in Inspection Procedure 71124.01.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

a. Inspection Scope

The inspectors evaluated whether the licensee controlled in-plant airborne radioactivity concentrations consistent with as low as reasonably achievable (ALARA) principles and that the use of respiratory protection devices did not pose an undue risk to the wearer.

During the inspection, the inspectors interviewed licensee personnel, walked down various portions of the plant, and reviewed licensee performance in the following areas:

The licensees use, when applicable, of ventilation systems as part of its engineering controls The licensees respiratory protection program for use, storage, maintenance, and quality assurance of NIOSH certified equipment, qualification and training of personnel, and user performance The licensees capability for refilling and transporting SCBA air bottles to and from the control room and operations support center during emergency conditions, status of SCBA staged and ready for use in the plant and associated surveillance records, and personnel qualification and training Audits, self-assessments, and corrective action documents related to in-plant airborne radioactivity control and mitigation since the last inspection These activities constitute completion of one sample of in-plant airborne radioactivity control and mitigation as defined in Inspection Procedure 71124.03.

b. Findings

No findings were identified.

OTHER ACTIVITIES

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

4OA1 Performance Indicator Verification

.1 Unplanned Scrams per 7000 Critical Hours (IE01)

a. Inspection Scope

The inspectors reviewed licensee event reports (LERs) for the period of January 2014 through December 2014 to determine the number of scrams that occurred. The inspectors compared the number of scrams reported in these LERs to the number reported for the performance indicator. Additionally, the inspectors sampled monthly operating logs to verify the number of critical hours during the period. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the data reported.

These activities constituted verification of the unplanned scrams per 7000 critical hours performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.2 Unplanned Power Changes per 7000 Critical Hours (IE03)

a. Inspection Scope

The inspectors reviewed operating logs, corrective action program records, and monthly operating reports for the period of January 2014 through December 2014 to determine the number of unplanned power changes that occurred. The inspectors compared the number of unplanned power changes documented to the number reported for the performance indicator. Additionally, the inspectors sampled monthly operating logs to verify the number of critical hours during the period. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the data reported.

These activities constituted verification of the unplanned power changes per 7000 critical hours performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.3 Unplanned Scrams with Complications (IE04)

a. Inspection Scope

The inspectors reviewed the licensees basis for including or excluding in this performance indicator each scram that occurred between January 2014 and December 2014. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the data reported.

These activities constituted verification of the unplanned scrams with complications performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.4 Occupational Exposure Control Effectiveness (OR01)

a. Inspection Scope

The inspectors reviewed corrective action program records documenting unplanned exposures or losses of radiological control over locked high radiation areas and very high radiation areas during the period of January through December, 2014. The inspectors reviewed a sample of radiologically controlled area exit transactions showing exposures greater than 100 mrem. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data.

These activities constituted verification of the occupational exposure control effectiveness performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.5 Radiological Effluent Technical Specifications (RETS)/Offsite Dose Calculation Manual

(ODCM) Radiological Effluent Occurrences (PR01)

a. Inspection Scope

The inspectors reviewed corrective action program records for liquid or gaseous effluent releases that occurred January through December, 2014, and were reported to the NRC to verify the performance indicator data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data.

These activities constituted verification of the radiological effluent technical specifications RETS/offsite dose calculation manual (ODCM) radiological effluent occurrences performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

.1 Routine Review

a. Inspection Scope

Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensees corrective action program and periodically attended the licensees condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified. The inspectors also reviewed the licensees problem identification and resolution activities during the performance of the other inspection activities documented in this report.

b. Findings

No findings were identified.

.2 Annual Follow-up of Selected Issues

a. Inspection Scope

The inspectors selected one issue for an in-depth follow-up:

On March 18, 2015, Condition Reports 29393 and 49529, refueling water storage tank aligned to non-safety related spent fuel pool cleanup system during Modes 1, 2, 3, and 4. The inspectors used Operating Experience Smart Sample 2012-02, Technical Specification Interpretation and Operability Determination, Revision 1 for assessing where a licensee credited compensatory measures, which substitute manual operator action for automatic action to perform a specified safety function, to consider/declare equipment operable.

The inspectors assessed the licensees problem identification threshold, cause analyses, extent of condition reviews and compensatory actions. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions were adequate to address the non-seismic spent fuel pool cleanup system aligned to the refueling water storage tank.

These activities constitute completion of one annual follow-up sample as defined in Inspection Procedure 71152.

b. Findings

No findings were identified.

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

.1 Temporary Diesel Generator Fire

a. Inspection Scope

On March 11, 2015, the inspectors were informed by the control room that a temporary diesel generator inside the protected area had an approximately two foot flame emitted from the exhaust stack. The inspectors responded to the site and monitored the licensees actions for the temporary diesel generator fire, reviewed station logs, and reviewed NUREG-1022, Event Reporting Guidelines, Revision 3, to ensure licensee compliance.

b. Findings

No findings were identified.

.2 Event Notification 50744 Retraction

a. Inspection Scope

On March 18, 2015, the licensee retracted Event Notification 50744 reported on January 19, 2015, that missile door 33012 protecting Class 1E engineered safety features, buses NB01/NB02 switchgear rooms was discovered misaligned on its hinge and stuck partially open. This was reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The inspectors reviewed the basis for the retraction and reviewed NUREG-1022, Event Reporting Guidelines, Revision 3, to ensure licensee compliance.

b. Findings

No findings were identified.

.3 (Closed) Licensee Event Report (LER) 05000482/2014-003-00: Failure of Safety

Injection Accumulator Vent Line Due to Low Stress - High Cycle Fatigue Results in Degraded Reactor Coolant Boundary

a. Inspection Scope

On April 20, 2014, at 10:30 a.m. during the stations Mid-Cycle Outage 20, a health physics technician observed water leaking approximately 2.5 gallons per hour from the 3/4-inch line upstream of safety injection system valve EPV0109. The leak was determined to be coming from a through-wall crack in the vent line for the combined safety injection and residual heat removal outlet piping to safety injection accumulator tank D.

The cause of the through-wall cracking was determined to be to low stress - high cycle fatigue. The same weld had experienced a previous failure. The evaluation of the November 2003 failure at this location had failed to include margin for vibrational impacts and variance in operational parameters resulting in inadequate corrective action to reduce vibration on the EPV0109 vent line.

The immediate corrective actions called for the flawed socket weld and vent valve assembly to be replaced on April 25, 2014. Dye penetrant examinations were performed in Mid-Cycle Outage 20 on similar unsupported socket weld vent/drain assemblies connected to ASME Code Class 1 piping with no indications identified.

The long term corrective action was to install a support on the EPV109 vent line during Refueling Outage 20 to reduce vibration.

On October 10, 2014, NRC Problem Identification and Resolution Inspection Report 05000482/2014007 (ML14283A612), documented NCV 05000482/2014-007-02, Failure to Preclude Repetition of a Significant Condition Adverse to Quality to Prevent Reactor Coolant System Leak related to this item.

This licensee event report was closed.

b. Findings

No findings were identified.

.4 (Closed) Licensee Event Report (LER) 05000482/2015-001-00: Personnel Error Causes

Two Inoperable Residual Heat Removal Trains

a. Inspection Scope

On January 28, 2015, the nightshift operations crew implemented a clearance order to support planned maintenance on residual heat removal valves EJHV8716A and EJHV8809A. At 5:34 a.m. on January 28, 2015, the oncoming crew identified that closing these valves rendered both trains of the emergency core cooling system to be inoperable. Operators entered Limiting Condition for Operation 3.0.3 and action was taken to restore valves EJHV8716A and EJH8809A to the open position.

The cause of the event was that licensed operators involved with the preparation and implementation of the clearance order did not recognize that current plant conditions could not support the proposed maintenance activity.

The licensee implemented the following corrective actions:

(1) Individuals involved with this event had their qualifications removed until remediation occurred;

(2) On January 29, 2015, the licensee issued Standing Order 36, Tagging Authority Duties, Revision 0, to provide specific guidance that affect equipment operability;

(2) On February 10, 2015, the electronic clearance order database was modified to identify the valves in Station Procedure AP 26C-004, Operability Determination and Functionality Assessment, Section A.16, that can cause entry into Limiting Condition for Operation 3.0.3;

(3) On February 19, 2015, Station Procedure AP 21D-003, Control of Tagging Information, was revised to identify the use of red switch boxes for the valves in AP26C-004 entry into Limiting Condition of Operation 3.0.3. The red switch boxes were placed on the control room boards in the control room to provide awareness to the operator of the significance of the valve.

b. Findings

.

One licensee identified finding was identified and documented in Section 4OA7 of this report.

These activities constitute completion of four event follow-up samples, as defined in Inspection Procedure 71153.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On March 13, 2015, the inspectors presented the radiation safety inspection results to Mr. C. Reasoner, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On March 19, 2015, the inspector conducted a telephonic exit meeting to present the results of the in-office inspection of changes to the licensees emergency plan to Mr. S. Koenig, Manager, Regulatory Affairs, and other members of the licensee staff. The licensee acknowledged the issues presented.

On March 26, 2015, the inspectors presented the In-Service Inspection team results to Mr. J.

McCoy, Vice President, Engineering, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspector had been returned or destroyed.

On April 1, 2015, the inspectors presented the inspection results to Mr. A. Heflin, President and Chief Executive Officer, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) 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 non-cited violation.

Technical Specification Section 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, February 1978. Section 1.c of Regulatory Guide 1.33 requires procedures for equipment control (e.g. locking and tagging). Station Procedure AP 21E-001, Clearance Orders, Revision 37, requires that the shift manager, ensure that plant conditions can support establishing the clearance order boundaries, including activities such as removing equipment from service. Contrary to the above, on January 28, 2015, the licensee failed to ensure that plant conditions could support the clearance order boundaries during preparation and implementation of clearance orders. Specifically, the preparation and implementation of clearance order EJ-A-005 unintentionally rendered both trains of the residual heat removal system inoperable and necessitated an unplanned entry into Technical Specification 3.0.3 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e. core damage). Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined a detail risk evaluation was required because this finding represented a loss of system and/or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst noted that the isolation of valve EJ HV8716A would only affect the reliability of hot leg injection for train B. Hot leg injection is a necessary function to ensure that there will not be unacceptably high concentrations of boric acid in the core region (resulting in precipitation of a solid phase) during the long-term cooling phase following a postulated large-break loss of coolant accident. Consequently, valve alignments affecting hot leg injection are only of concern during large-break loss of coolant accidents.

Using the simplified plant analysis risk model, the analyst noted that the frequency of a large-break loss of coolant accident (LLOCA) was 2.5 x 10-6 /year. As stated above, the exposure period was two hours or 2.28 x 10-4 years. The analyst then calculated the upper bound risk impact of the performance deficiency to be 5.7 x 10-10. Therefore, this finding is of very low safety significance (Green).

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

T. Baban, Manager, Systems Engineering

L. Bell, Engineer

P. Black, Human Resources

T. Branam, Design Engineer Electrical

L. Brinkley, Supervisor, Maintenance

A. Broyles, Manager, Information Services

T. Damashek, Simulator Fidelity Coordinator

P. Deblonk, Superintendent, Instrumentation and Control

B. Dorathy, Supervisor, Engineering Program

T. East, Supervisor, Emergency Planning

J. Edwards, Manager, Operations

D. Erbe, Manager, Security

M. Ferrel, Supervisor, Engineer

D. Ferrara, Supervisor, Quality

R. Fincher, Manager, Quality

R. Flannigan, Manager, Nuclear Engineering

J. Fritton, Owners Representative

L. Fure, Master Technician, Radiation Protection

A. Gilliam, ALARA Technician, Radiation Protection

D. Giefer, Engineer

A. Heflin, President and Chief Executive Officer

R. Hobby, Licensing

R. Jung, Instructor, Fire Protection

J. Knapp, Superintendent, Operations Training

S. Koenig, Manager, Regulatory Affairs

M. Legresley, System Engineer

D. Mand, Manager, Design Engineering

J. McCoy, Vice President Engineering

D. McDougal, Supervisor, Maintenance

C. Medenciy, Radiation Protection Supervisor

N. Mingle, System Engineer

W. Muilenburg, Supervisor, Licensing

J. Petty, System Engineer

E. Prather, Principal Engineer

E. Ray, Manager, Training

C. Reasoner, Site Vice President

B. Ryan, Licensed Operator Supervising Instructor

M. Skyles, Manager, Health Physics

S. Smith, Plant Manager

J. Steinert, Work Week Manager

A. Stull, Vice President and Chief Administrative Officer

K. Stuber, Supervisor, Maintenance

D. Sullivan, Manager, Supply Chain Services

J. Suter, Supervising Engineer, Fire Protection

B. Vickery, Manager, Financial Services

Attachment 1

D. Vu, PRA Engineer

S. Wideman, Licensing

J. Yunk, Manager, Corrective Actions

A. Yurko, Health Physics Technician

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000482/2015001- Question Related to Ultrasonic Examination of Reactor Vessel URI Flange Stud Hole Threads (1R08)

Opened and Closed

05000482/2015001- Failure to Assess the Operability of Emergency Diesel Generator NCV B during Emergent Work Activities (1R13)

Failure to Complete an Adequate Operability Evaluation for

05000482/2015001-

NCV Declaring the Train A Control Room Air Conditioning Unit

Operable (1R15.1)

05000482/2015001- Failure to Station Boundary Watch for Opening Auxiliary Building NCV Emergency Exhaust System Boundary Door (1R15.2)

Closed

Failure of Safety Injection Accumulator Vent Line Due to Low

05000482/2014-003-

LER Stress - High Cycle Fatigue Results in Degraded Reactor Coolant

Boundary (4OA3.3)

05000482/2015-001- Personnel Error Causes Two Inoperable Residual Heat Removal LER Trains (4OA3.4)