IR 05000482/2014002

IR 05000482-14-002; on 01/01/14 - 03/28/14; Wolf Creek Generating Station, Integrated Resident and Regional Report; Maintenance Effectiveness, Maintenance Risk Assessments and Emergent Work Control, and Operability Determinations and Functi

ML14133A153
Person / Time
Site:	Wolf Creek
Issue date:	05/12/2014
From:	O'Keefe N NRC/RGN-IV/DNMS/NMSB-B
To:	Heflin A Wolf Creek
References
IR-14-002
Download: ML14133A153 (49)

Text

May 12, 2014

SUBJECT:

WOLF CREEK GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000482/2014002

Dear Mr. Heflin:

On March 28, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Wolf Creek Generating Station. On April 9, 2014, the NRC inspectors discussed the results of this inspection with you and 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.

These findings involved violations of NRC requirements. Further, inspectors documented a licensee-identified violation 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 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/

Neil OKeefe, Chief Project Branch B Division of Reactor Projects Docket No.: 50-482 License No: NPF-42

Enclosure:

Inspection Report 05000482/2014002 w/ Attachments 1. Supplemental Information 2. TI 2515-182 Phase 2 Inspection Request 3. O

REGION IV==

Docket: 05000482 License: NPF-42 Report: 05000482/2014002 Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane NE Burlington, Kansas Dates: January 1 through March 28, 2014 Inspectors: C. Peabody, Senior Resident Inspector R. Stroble, Resident Inspector G. Callaway, Reactor Technology Instructor P. Hernandez, Health Physicist P. Jayroe, Reactor Inspector D. Proulx, Senior Project Engineer L. Ricketson, P.E., Senior Health Physicist D. You, Project Engineer Approved Neil OKeefe By: Chief, Project Branch B Division of Reactor Projects-1- Enclosure

SUMMARY

IR 05000482/2014002; 01/01/2014 - 03/28/2014; Wolf Creek Generating Station, Integrated

Resident and Regional Report; Maintenance Effectiveness, 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, 2014, 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. These findings involved violations of NRC requirements. Additionally, NRC inspectors documented in this report and a licensee-identified violation of very low safety significance (Green). 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, Components Within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRC's 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,

Procedures, for maintenance instructions inappropriate to the circumstances. Specifically,

Work Orders 11-341986-005 and 11-342065-002 did not contain adequate instructions for reassembling essential service water Garlock expansion joints to ensure proper joint alignment. As a result, on February 11, 2014, the inspectors identified that the inlet expansion joint for the essential service water intercooler heat exchanger, which provides cooling to emergency diesel generator B jacket water system, was misaligned by 0.5 inches, which exceeded the vendor specification of less than 0.125 inch. This item was entered into the corrective action program as Condition Reports 79352 and 79623, and the fitting was replaced during the mid-cycle 2014 outage. The licensee also conducted an extent of condition inspection and identified three additional Garlock expansion joints that were not made with the approved liner material.

The failure to properly reinstall essential service water expansion joints consistent with the vendor approved and analyzed configuration was a performance deficiency. The performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the misaligned Garlock expansion joint in the essential service water system degraded its long-term operability and its ability to withstand a seismic event.

Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time and the finding did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees

Maintenance Rule program for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Specifically, although the expansion joint was in a degraded condition, it was determined to be operable based on an engineering evaluation and seismic test data. The inspectors determined that the finding had a cross-cutting aspect in the human performance area of resources because the licensee did not ensure that personnel equipment, procedures, and other resources were available and adequate to support nuclear safety [H.1] (Section 1R12).

Green.

A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the failure to conduct excavation work such that it would ensure that design basis requirements for tornado missile protection and seismic qualification of safety-related cables were maintained during construction near the essential service water pump house. Specifically, when excavation near underground essential service water cables caused a loss of safety-related backfill over the cables, the licensee did not plan and execute the work in a manner that ensured that the qualified soil coverage around the train B essential service water duct bank was maintained by protecting against trench cave-ins.

Failure to maintain adequate soil coverage of the essential service water duct banks during construction is a performance deficiency. The deficiency is more than minor because it affected the protection against external factors and design control attributes of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Events Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. The inspectors determined that the finding had a cross-cutting aspect of work management in the area of human performance in that the process for planning, controlling, and executing work did not adequately include the identification and management of risk. Specifically, work planning did not account for adequate shoring material to prevent design basis ground cover from caving in during planned excavations in the vicinity of operable safety related equipment

[H.5] (Section 1R15).

Cornerstone: Barrier Integrity

Green.

A self-revealing non-cited violation, with two examples, of Technical Specification 5.4.1.a, Procedures, was identified for the failure to follow the reactivity management procedures. On two occasions, operators failed to take prudent actions to ensure that reactor power did not exceed the licensed limit of 3565 megawatts thermal while performing activities known to cause power increases.

On February 17, 2014, while performing chemical and volume control system inservice check valve testing on the discharge check valve of the train A centrifugal charging pump, operators performed a dilution of the reactor coolant system for normal power maintenance while reactivity was also being affected by the testing of the charging pump check valve, resulting in exceeding 100 percent power. On March 6, 2014, while returning the reactor to full power following data collection on the main turbine control valves, operators used an automatic power ramp to a setpoint of only 3 megawatts below 100 percent, without accounting for the overshoot that would result from the selected ramp rate, resulting in exceeding 100 percent power. In both cases, operators were alerted by an alarm indicating that the 1-minute average power level exceeded 100 percent. The inspectors reviewed station procedure GEN 00-004 Power Operation, and noted a requirement in Attachment A: For pre-planned evolutions that are expected to cause a transient rise in reactor power that could exceed the licensed power level, prudent actions should be taken to reduce power prior to the evolution.

Failure to take prudent action to maintain the reactor within licensed power limits prior to performing activities known to cause an increase in reactor power levels is a performance deficiency. The performance deficiency was more than minor because it affected both the configuration control attribute of reactivity control as well as the human performance attribute of procedure adherence of the Barrier Integrity Cornerstone, and impacted the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors screened the finding using the reactivity control screening questions found in Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Section C; question number 3 referred the inspectors to Inspection Manual Chapter 0609, Appendix M, Significance Determination Using Qualitative Criteria. NRC Management performed the qualitative assessment and determined that the finding was of very low safety significance (Green) because the relatively small magnitude of the overpower events, the prompt operator actions to return power to below the licensed limits upon discovery, and the fact that overpower events did not result in any failure of the fuel cladding. The inspectors determined that the finding had a conservative bias cross-cutting aspect in the area of human performance. Specifically, the affected evolutions were known in advance to have positive reactivity impacts; however, operators did not consider reducing power in the case of the check valve testing, nor was a slow approach to the maximum reactor power level used to avoid overshoot during dynamic turbine loading for the turbine valve data collection in order to prevent licensed power levels from being exceeded [H.14] (Section 1R13).

PLANT STATUS

Wolf Creek began the inspection period at 100 percent power. On March 5, 2014, power was reduced to 35 percent for turbine control valve testing and data collection. The next morning the unit returned to 100 percent power. On March 7, 2014, the unit was shutdown for a planned mid-cycle outage. 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

.1 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On February 10, 2014, the inspectors completed an inspection of the stations readiness for impending adverse weather conditions. The inspectors reviewed circulating water screen house fire pump features, the licensees procedures to respond to prolonged freezing conditions, ice and snow, and the licensees planned 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

.1 Partial Walkdown

a. Inspection Scope

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

• January 7, 2014, motor driven auxiliary feedwater train A

• February 19, 2014, component cooling water train B

• February 25, 2014, centrifugal charging pump B while centrifugal charging pump A was inoperable for planned maintenance 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 three partial system walk-down samples as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

.2 Complete Walkdown

a. Inspection Scope

On February 26, 2014, the inspectors performed a complete system walk-down inspection of the essential service water system. The inspectors reviewed the licensees procedures and system design information to determine the correct system lineup for the existing plant configuration. The inspectors then visually verified that the system was correctly aligned for the existing plant configuration.

These activities constituted one complete system walk-down sample, as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 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 five plant areas important to safety:

• January 8, 2014, nonvital switchgear/battery rooms, fire areas C-17 and C-18

• March 11, 2014, train A emergency diesel generator room, fire area D-1

• March 13, 2014, train A residual heat removal pump room, fire area A-2

• March 13, 2014, train B residual heat removal pump room, fire area A-4

• March 13, 2014, train A and train B component cooling water system areas, fire area A-16 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 five quarterly inspection samples, as defined in Inspection Procedure 71111.05.

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 February 25, 2014, the inspectors observed simulator training for an operating crew.

The inspectors assessed the performance of the operators and the evaluators critique of their performance.

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

On March 7 and 8, 2014, 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 and risk due to planned reactor shutdown for midcycle maintenance outage. The inspectors observed the operators performance of the following activities:

• Activity 1, shift crew brief

• Activity 2, transition from main feed regulating valves to bypass feed regulating valves

• Activity 3, boration and control and shutdown rod insertion

• Activity 4, turbine trip and coast down In addition, the inspectors assessed the operators adherence to plant procedures, including AP 21-001, Revision 67, Conduct of Operations, 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 an instance of degraded performance or condition of safety-related structures, systems, and components (SSCs):

• February 10, 2014, degraded Garlock expansion joints 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

Introduction.

The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, Procedures, for failure to properly re-install Garlock expansion joints on the essential service water system.

Description.

On February 10, 2014, while touring the emergency diesel generator rooms, the inspectors noted that several of the expansion joints for the essential service water supply to the jacket water intercooler heat exchangers for the emergency diesel generators had cracking and tearing on the covers. In addition, the essential service water intercooler inlet (EKJ03B) expansion joint for emergency diesel generator B appeared to be misaligned in the lateral direction. The inspectors informed licensee personnel of this observation, and the licensee initiated Condition Reports 79352 and 79623 to place this item into the corrective action program.

The inspectors obtained a copy of the vendor manual for the expansion joints, titled Garlock Sealing Technologies, Expansion Joints Technical Manual. The installed expansion joints on the essential service water system were 6 inch, Style 206, EZ Flow, expansion joints designed for low pressure applications (250 psig). The expansion joints were made of a rubber cover and a nylon reinforced core. The installation instructions directed the user to check the cover for damage and to ensure that the joint was installed with a maximum lateral misalignment of 0.125 inches. The troubleshooting section of the vendor manual stated that if there was cracking at the base of the arch or flange, to check for proper alignment (i.e., less than 0.125 inches).

In addition, the inspectors obtained the Electric Power Research Institute (EPRI) industry guidance for flexible joints, Expansion Joint Maintenance Guide, Revision 1, for further standard practice. Section 4.1.2 provided directions on inspection of expansion joints and recommended that the outer cover of the expansion joints be examined for cracks and tears, ensuring that none of the nylon from the inner core was exposed. Cracks that could allow moisture intrusion into the joint should be repaired. In addition, the EPRI guide recommended that the joint be checked for lateral and angular misalignment.

The licensee measured the total misalignment of the B emergency diesel generator essential service water intercooler and determined that the lateral misalignment was approximately 0.5 inches, in excess of the analyzed static misalignment of 0.125 inches.

The licensee performed an operability evaluation that demonstrated that the maximum overall lateral deflection during a seismic event (taken into account the initial misalignment) would be 0.538 inches, less than the vendor recommended 0.625 inches, and thus was an operable, but degraded condition. Also, the licensee noted that the cracks and tears noted on the covers did not expose the inner core and thus did not affect the operability of the expansion joints.

The licensee determined that the essential service water to jacket water intercooler expansion joints had been replaced in 2008. The applicable installation procedures had sufficient detailed instructions that verified proper condition and alignment including quality control verification signatures. The licensee also allowed the essential service water expansion joints to be disassembled to support outage maintenance. The essential service water to intercooler inlet expansion joints had last been disassembled in February 2012 to support retubing of the emergency diesel generators intercooler heat exchanger tubes using Work Orders 11-341986-005 and 11-342065-002. These work orders did not contain sufficient directions to ensure proper alignment of the expansion joints and were likely the cause of the misalignment. Thus, Work Orders 11-341986-005 and 11-342065-002 were not appropriate to the circumstances.

The inspectors reviewed completed periodic inspections associated with the essential service water expansion joints. The most recent performance of these was in Work Order 10-325817-000 conducted in May 2010. The inspectors noted that this work order only contained a listing of the affected components and a signature block indicating completion of the inspection. No acceptance criteria or any other directions were provided. The user only entered the words Sat (for satisfactory) in the comments section of the work order. The licensee revised the work order for the next inspection of expansion joints, to be performed during the spring 2015 refueling outage, to include the appropriate recommended acceptance criteria.

Because of these observations, the licensee conducted an extent of condition review and determined that inspection and installation procedures, and work orders for other expansion joints, did not contain adequate instructions to ensure proper condition and alignment. The licensee stated that they would revise these procedures. Additionally, the licensee contacted the vendor concerning the condition/misalignment of the emergency diesel generator B intercooler inlet expansion joint. The licensee replaced this expansion joint during the spring 2014 mid-cycle outage.

Analysis.

The failure to properly reinstall essential service water expansion joints consistent with the vendor-approved and analyzed configuration was a performance deficiency. The performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the misaligned Garlock expansion joint in the essential service water system degraded its longterm operability and its ability to withstand a seismic event. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time and the finding did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Specifically, although the expansion joint was degraded, it was determined to be operable based on an engineering evaluation and seismic test data. The inspectors determined that the finding had a cross-cutting aspect in the human performance area of resources because the licensee did not ensure that personnel equipment, procedures, and other resources were available and adequate to support nuclear safety [H.1].

Enforcement.

Technical Specification 5.4.1.a states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 9.a, Procedures for Performing Maintenance, states, in part, that 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 to the circumstances. Contrary to the above, maintenance that affected a safety-related system (essential service water system) was not performed with procedures or documented instructions appropriate to the circumstances. Specifically, Work Orders 11-341986-005 and 11-342065-002, implemented in February 2012, did not contain appropriate instructions to ensure lateral misalignment of less than 0.125 inches for essential service water expansion joints. As a result, as of February 11, 2014, the essential service water inlet expansion joint for the emergency diesel generator B intercooler heat exchanger was degraded and misaligned approximately 0.5 inches. Because the violation was of very low safety significance (Green) and was entered into the licensees corrective action program as Condition Reports 79352 and 79623, this is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000482/2014002-01, Inadequate Work Instructions for Reinstallation of Essential Service Water Expansion Joints.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed four 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:

• February 10, 2014, reactor coolant system loop 4 flow instrument failure

• March 5 and 6, 2014, moisture separator reheater testing and turbine control system testing

• March 12, 2014, daily outage risk assessment

• March 14, 2014, reactor coolant system pressure transient 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.

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 SSCs.

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

b. Findings

Introduction.

A Green self-revealing non-cited violation, with two examples, of Technical Specification 5.4.1.a, Procedures, was identified for the failure to follow reactivitiy management procedures. On two occasions, operators failed to take prudent actions to ensure that reactor power did not exceed the licensed limit of 3565 megawatts thermal while performing activities known to cause power increases.

Description.

Example 1 On February 17, 2014, while performing chemical and volume control system inservice check valve testing on the discharge check valve of the train A centrifugal charging pump, operators received an alarm indicating that the reactor had exceeded the licensed thermal power limit as indicated by the 10 minute calorimetric moving average.

Operators took immediate action to restore power by injecting soluble boron into the reactor coolant system and inserting control rods.

The inspectors reviewed Condition Report 79701 and interviewed operations personnel regarding the incident. The inspectors noted that operators had continued to make routine power maintenance dilutions throughout this evolution to keep the reactor at full power. The inspectors asked the operators why this evolution, which was known to add positive core reactivity, would be performed at 100 percent power, and the licensee acknowledged that a procedure change to reduce power to gain margin to the licensed thermal power limit during this type of testing was under consideration in the corrective action program, but this was not acted upon because the power reduction was not specified by the procedure.

The inspectors also questioned whether performing power maintenance dilutions was appropriate during evolutions known to affect reactivity with no additional margin to the licensed thermal power limits, and the operators acknowledged that while not explicitly prohibited by station procedures, it was not a good practice. The inspectors reviewed the plant computer data and found that the plant had operated above the licensed power limit for approximately 6 minutes, and that the maximum power reached was 3569 megawatts thermal, 4 megawatts greater than the licensed power limit of 3565 megawatts thermal.

Example 2 On March 6, 2014, while returning the reactor to full power following data collection on the main turbine control valves, operators received an alarm indicating that the reactor had exceeded the licensed thermal power limit as indicated by the 10 minute calorimetric moving average. Operators took immediate action to restore power by injecting soluble boron into the reactor coolant system and inserting control rods.

The inspectors reviewed Condition Report 80353 and interviewed operations personnel regarding the incident. In this case, the procedure had specified using a fixed turbine loading ramp rate back to full power. Operators set the turbine controller to stop at 3 megawatts below 100 percent power. However, when the turbine load increase stopped, the reactor transient overshot the licensed thermal power limit. Operators did not use a slow, cautious approach to 100 percent power and were not closely monitoring power levels as the limit was approached. The inspectors questioned the operators why this evolution, which was known previously to impact core reactivity, would be allowed to exceed 100 percent power without proper operator oversight as discussed in the pre-job reactivity brief and the licensee acknowledged that power levels should have been monitored more closely. The inspectors reviewed the plant computer data and found that the plant had operated above the licensed power limit for approximately 5 minutes, and that the maximum power reached was 3570 megawatts thermal, 5 megawatts greater than the licensed power limit of 3565 megawatts thermal.

The inspectors also reviewed station, regulatory, and industry guidance regarding compliance with licensed reactor thermal power limits, outlined in Regulatory Information Summary 2007-21, Revision 1, Adherence to Licensed Power Limits, and its endorsed Nuclear Energy Institute (NEI) Position Statement entitled Guidance to Licensees on Complying with the Licensed Power Limit, as well as the corresponding NRR Safety Evaluation. The inspectors concluded that the guidance of Section 4.3(2) was not met.

Specifically, for evolutions expected to cause a transient increase in reactor power that could exceed the licensed power limit value, prudent action should be taken beforehand to ensure that the licensed power limit is not exceeded. The inspectors reviewed station procedures AP 21-001, Conduct of Operations, and GEN 00-004, Power Operations, and found that Attachment A of GEN 00-004 contains the NEI position statement.

The licensee did not agree that these events were not consistent with the NRC-endorsed guidance. Specifically, the licensee pointed to the parts of the guidance that allowed power that was experiencing normal fluctuations to exceed 100 percent provided operators reduced power such that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> power average did not exceeded 100 percent. The staff concluded that this guidance did not apply when operator actions caused the power to depart from steady state conditions; during a transient caused by operator action, normal fluctuations are not being experienced. Specifically, the relevant guidance supported the NRC staff conclusion:

The Wolf Creek operating license permits operation at reactor power levels not in excess of 3565 megawatts thermal.

RIS 2007-21, Revision 1, states that licensees may not intentionally operate or authorize operation above the maximum power level as specified in the license.

This document endorsed the NEI Position Statement providing Guidance to Licensees on Complying with the Licensed Power Limit.

The NRC-endorsed guidance defines the term steady state to involve temperatures, pressures, and flows are stable such that the nominal value of reactor power remains stable, subject to statistical uncertainties and normal fluctuations. Also, small, short-term fluctuations in power that are not under the control of a licensed reactor operator are not considered intentional. This document included examples of performance deficiencies; example 4 included the failure to take prudent action prior to a pre-planned evolution that could cause a power increase to exceed the licensed power level.

The staff also noted that the NRC accepted verifying core thermal power by using an average value derived from secondary plant indications for steady-state conditions.

However, the NRC safety evaluation for RIS 2007-21, Revision 1, states that power needs to be monitored on a frequency that ensures the parameter has not exceeded the limit and that an adverse trend can be detected in a timely manner.

Analysis.

Failure to take prudent action to maintain licensed power limits prior to performing maintenance activities known to cause an increase in reactor power levels is a performance deficiency. The performance deficiency is more than minor because it affected both the configuration control attribute of reactivity control as well as the human performance attribute of procedure adherence of the Barrier Integrity Cornerstone, and impacted the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors screened the finding using the reactivity control screening questions found in Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Section C, question number 3 referred the inspectors to Inspection Manual Chapter 0609, Appendix M, Significance Determination Using Qualitative Criteria. NRC Management performed the qualitative assessment and determined that the finding was of very low (Green) safety significance because the relatively small magnitude of the overpower, the prompt operator actions to return power to below the licensed limits upon discovery, and the fact that overpower events did not result in any failure of the fuel cladding (fuel leaks into the reactor coolant system). The inspectors determined that the finding had a conservative bias cross-cutting aspect in the area of human performance. Specifically, the affected maintenance evolutions were known in advance to have positive reactivity impacts; however, operators did not consider reducing power in the case of the check valve testing, nor was a slow approach to the maximum reactor power level used to avoid overshoot during dynamic turbine loading for the turbine valve data collection in order to prevent licensed power levels from being exceeded [H.14].

Enforcement.

Wolf Creek Technical Specification 5.4.1.a states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, Appendix A, February, 1978. Regulatory Guide 1.33, Appendix A, Section 2.f and 2.g, requires general operating procedures for changing load and power operation, respectively. Station procedure GEN 00-004, Power Operations, Revision 77, implemented these requirements. GEN 00-004, Appendix A, Section A.1 final bullet required that for pre-planned evolutions expected to cause a transient rise in reactor power that could exceed the Licensed Power Level, prudent action should be taken to reduce power prior to the evolution. Contrary to the above, on two occasions, operators failed to implement procedures required for load changing and power operation. Specifically, on February 17, 2014, during chemical and volume control system check valve testing, a planned maintenance evolution known to affect core reactivity, operator action caused the reactor to be operated above rated thermal power for approximately six minutes at 3669 megawatts thermal, a non-implementation of the guidelines specified in GEN 00-004, Attachment A. Also contrary to the above, on March 6, 2014, during main turbine valve data collection, a planned maintenance evolution known to affect core reactivity, operator action caused the reactor to be operated above rated thermal power for approximately five minutes at 3570 megawatts thermal a non-implementation of the guidelines specified in GEN 00-004, Attachment A. Because the violation was of very low safety significance (Green) and was entered into the licensees corrective action program as Condition Reports 79701 and 80353, respectively, this is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000482/2014002-02, Failure to Maintain Licensed Power Limits During Planned Evolutions Affecting Reactivity.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors reviewed three operability determinations and functionality assessments that the licensee performed for degraded or nonconforming SSCs:

• January 8, 2014, train B main feedwater pump hydraulic actuator

• January 20, 2014, essential service water intake structure cave-in

• February 16, 2014, emergency diesel generator Garlock gasket misalignment, OE KJ-14-002 The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded SSC to be operable or functional, the inspectors verified that the licensees compensatory measures were appropriate to provide reasonable assurance of operability or functionality. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability or functionality of the degraded SSC.

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

b. Findings

Introduction.

A Green self-revealing, non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the failure to ensure that design basis requirements for tornado missile protection and seismic response of safety-related cables were maintained during construction near the essential service water pump house. Specifically, the licensee did not ensure that the qualified soil coverage around the train B essential service water duct bank was maintained.

Description.

On January 20, 2014, while excavating material on the north side of the essential service water pump house, a portion of the excavation wall caved in, reducing the seismic and missile protection qualified soil coverage of train B essential service water electrical duct bank below the required 4.5 feet minimum to an estimated 1 foot of coverage, an insufficient amount to retain design margins. The material washout was estimated to be about 180 cubic yards. The licensee had planned to maintain train B essential service water operable during this work.

Despite digging in close proximity to a buried essential service water cable duct bank while that train was expected to be operable, the work planners failed to recognize the potential for the trench walls to collapse and either expose the duct bank or reduce the safety-related backfill required for tornado missile protection and seismic response.

Extensive trenching was being done to support the installation of new essential service water piping. Had the potential for trench wall collapse been recognized, the licensee could have installed shoring material to prevent the trench from caving in. This was determined to be self-revealing because very little analysis was needed to recognize the cave-in or the impact to the essential service water system. This issue was entered into the licensees corrective action program as Condition Report 78664.

The licensee identified a related violation that resulted from the licensees initial response to the cave-in. A team was dispatched to verify locally the design margins, and when they reported back to the control room, the decision was made to refill the hole as soon as possible. The licensee promptly ordered 150 cubic yards of rip-rap to be used to recover the duct bank. Unlike compacted soil, rip-rap is coarse rock typically about the size of a football that is used to prevent erosion on shorelines or embankments. While it does provide some missile protection, it behaves very differently during seismic events and does not meet the requirements of the design bases. The material deposition was completed at 3:38 a.m. on January 21, 2014.

The following day shift, a senior reactor operator and design engineer both came to the conclusion that the rip-rap was not properly qualified. The essential service water train was declared operable, but degraded, and the material would have to be removed and replaced during the spring 2014 mid-cycle outage. The deposition of rip-rap in place of a qualified backfill material was a licensee-identified violation, which is documented in Section 4OA7 of this report.

Analysis.

Failure to maintain adequate soil coverage of the essential service water duct banks during construction is a performance deficiency. The performance deficiency is more than minor because it affected the protection against external factors and design control attributes of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Events Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. The inspectors determined that the finding had a cross-cutting aspect of work management in the area of human performance in that the process for planning, controlling, and executing work did not adequately include the identification and management of risk. Specifically, work planning did not account for adequate shoring material to prevent design basis ground cover from caving in during planned excavations in the vicinity of operable safety-related equipment [H.5].

Enforcement.

10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled.

Measures shall also 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 January 20, 2014, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, work instructions for conducting excavations near the train B essential service water cable duct bank did not sufficiently shore up the foundation of the essential service water intake structure such that the underground duct banks maintained adequate seismic and missile protection in accordance with applicable design basis requirements. Because the violation was of very low safety significance and was entered into the licensees corrective action program as Condition Report 78664, it is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000482/2014002-03, Failure to Maintain Seismic and Missile Protection Design Basis Requirements During Essential Service Water Construction.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed three post-maintenance testing activities that affected risk-significant SSCs:

• January 22, 2014, safety injection pump A lube oil sample and analysis

• February 27, 2014, centrifugal charging pump A post-maintenance test

• March 13, 2014, safety injection pump B recirculation to refueling water storage tank valve post-maintenance test 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 three 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 planned mid-cycle outage that started on March 8, 2014, and continued throughout the remainder of the inspection period, 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 These activities constitute completion of one planned outage activities sample, as defined in Inspection Procedure 71111.20.

b. Findings

No findings were identified

1R22 Surveillance Testing

a. Inspection Scope

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

ASME In-service tests:

• January 24, 2014, essential service water train A check valve operability, EF-V046 Other surveillance tests:

• January 9, 2014, pressurizer level channel calibration

• January 21, 2014, containment seismic monitor channel operational test

• February 27, 2014, slave relay test (K640) train A motor driven auxiliary feedwater pump start 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 tests satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected SSCs following testing.

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

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.

b. Findings

No findings were identified.

2RS2 Occupational ALARA Planning and Controls

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining occupational individual and collective radiation exposures as low as is reasonably achievable (ALARA). During the inspection, the inspectors interviewed licensee personnel and reviewed licensee performance in the following areas:

• Site-specific ALARA procedures and collective exposure history, including the current 3-year rolling average, site-specific trends in collective exposures, and source-term measurements

• ALARA work activity evaluations/postjob reviews, exposure estimates, and exposure mitigation requirements

• The methodology for estimating work activity exposures, the intended dose outcome, the accuracy of dose rate and man-hour estimates, and intended versus actual work activity doses and the reasons for any inconsistencies

• Records detailing the historical trends and current status of tracked plant source terms and contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry

• Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas

• Audits, self-assessments, and corrective action documents related to ALARA planning and controls since the last inspection These activities constitute completion of one sample of occupational ALARA planning and controls as defined in Inspection Procedure 71124.02.

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 2013 through December 2013 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 2013 through December 2013 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 Outages 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 2013 and December 2013. 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 2013. 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 2013 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 daily screening review team 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 two issues for an in-depth follow-up:

• On February 27, 2014, the inspectors reviewed a sample of corrective actions assigned to the operations training department to verify that problems being entered into the corrective action program were being entered at the appropriate threshold and actions to address resolution were appropriate and timely.

• Multiple failures of 480V molded case circuit breakers In December of 2013, the inspectors observed a possible trend in the area of breaker failures during daily reviews of condition reports. Upon communicating this information to the licensee, they performed a basic trend analysis under Condition Report 77730.

These activities constitute completion of two annual follow-up samples, as defined in Inspection Procedure 71152.

b.

Observations and Assessments For the first sample, the inspectors observed that actions assigned to operations training were appropriately prioritized; higher-priority condition reports, apparent and root cause evaluations correctly determined the cause of the problem and directed appropriate actions to preclude repetition. The inspectors determined that operations training assigned condition reports were being appropriately trended at the department and station level.

Following a review of the molded case circuit breaker failures, the inspectors concluded that a majority of the failures were part of the licensee monitoring programs to replace components prior to a demand failure. No significant trend in breaker failures was apparent. The inspectors concluded that the licensee was initiating condition reports at the appropriate threshold. Furthermore, cause analyses and extent of condition reviews and compensatory actions were appropriate to the circumstance. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions were adequate to correct the conditions identified.

c. Findings

No findings were identified.

4OA5 Other Activities

.1 (Closed) Temporary Instruction 2515/182 - Review of the Industry Initiative to Control

Degradation of Underground Piping and Tanks

a. Inspection Scope

Leakage from buried and underground pipes has resulted in groundwater contamination incidents with associated heightened NRC and public interest. The industry issued a guidance document, NEI 09-14, Guideline for the Management of Buried Piping Integrity, (ADAMS Accession No. ML1030901420) to describe the goals and required actions (commitments made by the licensee) resulting from this underground piping and tank initiative. On December 31, 2010, NEI issued Revision 1 to NEI 09-14, Guidance for the Management of Underground Piping and Tank Integrity, (ADAMS Accession No. ML110700122) with an expanded scope of components which included underground piping that was not in direct contact with the soil and underground tanks. On November 17, 2011, the NRC issued Temporary Instruction 2515/182, Review of the Industry Initiative to Control Degradation of Underground Piping and Tanks, to gather information related to the industrys implementation of this initiative.

b. Observations The licensees buried piping and underground piping and tanks program was inspected in accordance with paragraph 03.02.a of the temporary instruction and it was confirmed that activities which correspond to completion dates specified in the program which have passed since the Phase 1 inspection was conducted have been completed. Additionally, the licensees buried piping and underground piping and tanks program was inspected in accordance with paragraph 03.02.b of the temporary instruction and responses to specific questions were submitted to the NRC headquarters staff. Based upon the scope of the review described above, Phase II of TI-2515/182 was completed.

c. Findings

No findings were identified.

.2 (Closed) Violation 05000482/2013003-08 - Failure to Maintain Complete and Accurate

Housekeeping Records (EA-13-084)

This violation of 10 CFR 50.9 for failure to provide complete and accurate information regarding housekeeping observation records which are required by the stations Fire Protection Program was issued in the Second Quarter 2013 Wolf Creek Resident Inspectors Integrated Inspection Report. The inspectors reviewed the licensees corrective actions under Condition Report 71457 regarding this violation, including the licensees docketed response, cause evaluation, observation tracking records, security card reader cross-checks, interim effectiveness reports, revisions to the Wolf Creek Nuclear Operating Corporations (WCNOC) Code of Ethics station procedure AI 13C-002, and the WCNOC Principles of Integrity, a corporate publication distributed to employees. The licensees evaluation determined that the root cause to be the individuals decision to falsify records and subsequently deny doing so. However, the contributing cause of insufficient management oversight appears to have been addressed by the corrective action plan. Furthermore, the inspectors are aware of the results of licensee management verifying the performance of activities related to quality using security identification card reader logs, and when wrongdoing is apparent or suspected the licensee has brought the matter to the attention of the NRC Staff.

VIO 05000482/2013003-08 (EA-13-084) is closed.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On April 9, 2014, the inspectors presented the resident inspection results to Mr. A. Heflin, 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 21, 2014, the inspectors presented the radiation safety inspection results to Mr. R.

Smith, 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 January 23, 2014, the inspector presented the results of the underground piping inspection to Mr. R. Smith, 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 inspector 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 meet the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation.

• A licensee-identified violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for failure to ensure that design basis requirements were maintained in response to a cave-in of required essential service water ground cover. Specifically, the licensee did not ensure that the qualified soil coverage around the train B essential service water duct bank was maintained when they re-covered the duct banks with an unapproved material. Contrary to these requirements, on January 20, 2014, upon a loss of essential service water duct bank soil coverage due to a cave-in, the licensee refilled the voided area with an unapproved material that was not qualified to withstand seismic and missile design basis accidents.

The performance deficiency was failure to ensure the appropriateness of seismic and missile-qualified material. This violation was more than minor because it affected the protection against external factors and design control attributes of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Events Screening Questions, the inspectors determined that the finding was of very low safety significance because it did not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. Since the finding was licensee-identified, no cross-cutting aspect is assessed. The finding was entered into the licensees corrective action program as Condition Report 79089.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

R. Ayers, Supervisor, Radiation Protection

T. Baban, Manager Systems Engineering

P. Bedgood, Manager, Radiation Protection

M. Blow, Shift Manager

M. Brinkmeyer, Fire Protection Technician

A. Broyles, Manager Information Services

A. Camp, Plant Manager

D. Campbell, Superintendent Electrical Maintenance

R. Clemens, Vice President Strategic Projects

T. East, Superintendent E-Plan

D. Erbe, Manager Security

B. Evans, Training

R. Flannigan, Manager Nuclear Engineering

K. Fredrickson, Licensing

R. French, Radiation Protection Supervisor

A. Heflin, President and CEO

S. Henry, Vice President Engineering

R. Hobby, Licensing Engineer

T. Jensen, Manager Chemistry

A. Keneipp, Supervisor Quality

S. Koenig, Manager Strategic Initiatives

N. Mayhew, Program Engineer

M. McMullen, Design Engineer

C. Medenciy, Radiation Protection Supervisor

P. Moore, Operations

W. Muilenburg, Supervisor Licensing

E. Peterson, Ombudsman

L. Ratzlaff, Manager Maintenance

L. Sawyer, Supervisor Corrective Action

M. Skiles, Radiation Protection Supervisor

R. Skiles, Senior Technical Trainer

T. Slenker, Operations CAPCO

R. Smith, Site Vice President

M. Westman, Manager Regulatory Affairs

S. Wideman, Licensing Engineer

J. Wilson, Program Engineer

J. Yunk, Manager Corrective Actions

A. Yurko, Health Physics Technician

NRC Personnel

T. Kolb, NRR

A. Lewin, NRR

Attachment 1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000482/2014002- NCV Inadequate Work Instructions for Reinstallation of ESW Expansion Joints (Section 1R12)

05000482/2014002- NCV Failure to Maintain Licensed Power Limits During Planned Evolutions Affecting Reactivity (Section 1R13)

05000482/2014002- NCV Failure to Maintain Seismic and Missile Protection Design Basis Requirements During Essential Service Water Construction (Section 1R15)

Closed

Temporary TI Review of the Industry Initiative to Control Degradation of Instruction 2515/182 Underground Piping and Tanks (Section 4OA5)

05000482/2013003- VIO Failure to Maintain Complete and Accurate Housekeeping Records (EA-13-084) (Section 4OA5)