IR 05000482/2010002

IR 05000482-10-002; 01/01/2010 - 03/31/2010; Wolf Creek Generating Station; Integrated Resident and Regional Report; Fire Protection, Maintenance Effectiveness, Surveillance Testing, Event Response

ML101241075
Person / Time
Site:	Wolf Creek
Issue date:	05/04/2010
From:	Geoffrey Miller NRC/RGN-IV/DRP/RPB-B
To:	Matthew Sunseri Wolf Creek
References
IR-10-002
Download: ML101241075 (40)

Text

May 4, 2010

SUBJECT:

WOLF CREEK GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000482/2010002

Dear Mr. Sunseri:

On March 31, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Wolf Creek Generating Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on April 8, 2010, with you and other members of your staff.

The inspections examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents five NRC identified findings of very low safety significance (Green). Four of these findings were determined to involve violations of NRC requirements. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as a noncited violations, consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest the violations or the significance of the noncited violations, 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, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Wolf Creek Generating Station.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its enclosure, will be available electronically for public inspection in the NRC Public Document

Room or from the Publicly Available Records component of NRCs document 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/

Geoffrey B. Miller, Chief Project Branch B

Division of Reactor Projects

Docket No. 50-482 License No. NPF-42 Enclosure Inspection Report 05000482/2010002

w/Attachment:

Supplemental Information

REGION IV==

Docket:

05000482 License:

NPF-42 Report:

05000482/2010002 Licensee:

Wolf Creek Nuclear Operating Corporation Facility:

Wolf Creek Generating Station Location:

1550 Oxen Lane SE Burlington, Kansas Dates:

January 1 through March 31, 2010 Inspectors:

C. Long, Senior Resident Inspector C. Peabody, Resident Inspector J. Melfi, Project Engineer J. Groom, Resident Inspector, Callaway D. Dumbacher, Senior Resident Inspector, Callaway G. Guerra, CHP, Emergency Preparedness Inspector Approved By:

G. Miller, Chief, Project Branch B Division of Reactor Projects

Enclosure

- 1 -

SUMMARY OF FINDINGS

IR 05000482/2010002, 1/01/2010 - 3/31/2010; Wolf Creek Generating Station, Integrated

Resident and Regional Report; Fire Protection, Maintenance Effectiveness, Surveillance Testing, Event Response The report covered a 3-month period of inspection by resident inspectors and one announced baseline inspection by regional based inspectors. Five Green findings, four of which were noncited violations, were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination Process. Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of License Condition 2.C(5)(a) for degraded fire seals that separated redundant safe shutdown equipment. Specifically, silicone foam and ceramic fiber board seals separating the auxiliary feedwater trains from the turbine building and the condensate storage tank valve house were degraded so that they no longer provided a 3-hour rated fire barrier. The licensee entered the finding into the corrective action program as Condition Report 23828.

The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external events and affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix F,

Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, under Fire Barrier Degradation, Table A2.2, the finding was associated with Moderate B degradation due to the seal not being in a tested or evaluated condition.

Using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, in supplemental screening for fire confinement findings, the finding screens as Green due to exposing Fire Area A33 featuring an automatic full area water-based suppression system. No crosscutting aspect was assigned as this condition was not reflective of current licensee performance (Section 1R05).

Green.

The inspectors identified a noncited violation of 10 CFR 50.65 for failure to establish goals per paragraph (a)(1) to monitor the performance of the main condenser offgas radiation Monitor GERE0092. Multiple failures occurred which exceeded the monitoring goals and the function was not moved to 50.65(a)(1) status for corrective action and goal setting. Wolf Creek engineering subsequently evaluated the issues and determined that the function should have been moved to a(1) for goal setting. The licensee entered this issue in their corrective action program as Condition Report 24423.

This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events. The inspectors evaluated the significance of this finding using Inspection

Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," and determined that this finding is of very low safety significance,

Green.

Specifically, the associated function (SP-04) to detect primary to secondary leakage and then isolate the steam generator blowdown flow path does not result in a loss of any safety function.

The inspectors determined that this finding has a crosscutting aspect in the problem identification and resolution area associated with corrective action program because Wolf Creek failed to take appropriate corrective actions to address the system reliability issue and adverse radiation monitor performance trends in a timely manner, commensurate with safety significance and complexity P.1(d) (Section 1R12).

Green.

The inspectors identified a Green finding for the failure to adequately implement the posttrip review procedure following a reactor trip caused by low steam generator water levels on March 2, 2010. Specifically, Wolf Creek's posttrip evaluation was not adequate because it failed to identify or evaluate anomalous equipment performance associated with the main feedwater pump that caused the trip. Additionally, the inspectors determined that the Wolf Creeks posttrip review failed to identify that some aspects of operator response to the trip of the main feedwater pump were not in accordance with station procedures. Wolf Creek evaluated the individual issues and deficiencies listed above and entered them into the corrective action program as Condition Reports 23932, 23966, 24043, 23982, and 23981.

This finding was greater than minor because the information omitted from the posttrip review was associated with the human performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since the finding does not represent a loss of system safety function, nor does the finding represent actual loss of safety function for single train for a greater time than permitted by technical specifications. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because Wolf Creek failed to fully evaluate plant computer data and operator statements associated with the March 2, 2010, reactor trip P.1(c) (Section 4OA3).

Green.

The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a, Procedures, for the failure of Wolf Creek control room personnel to follow procedures for a main feedwater pump trip. During a review of the posttrip data and operator statements, the inspectors noted that control room operators took manual control and reset main feedwater Pump A, which was not in accordance with station procedures. This issue was entered into the licensee's corrective action program as Condition Report 24011.

This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of human performance and it affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609.04,

Phase 1 - Initial Screening and Characterization of Findings, and screened the finding to Phase 2 because the finding represents a loss of auxiliary feedwater actuation system safety Function g. The finding screened to Phase 3 because of the failure to start of both motor-driven auxiliary feedwater pumps. The senior reactor analyst performed a Phase 3 analysis and concluded that the finding was Green because the probability of an initiator occurring within any 10-second exposure time is approximately 3E-7.

Additionally, auxiliary feedwater pumps would have been automatically started on lo-lo steam generator level if required. The inspectors also determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with operating experience because Wolf Creek failed to communicate relevant operating experience to affected internal stakeholders. P.2(a) (Section 4OA3).

Cornerstone: Barrier Integrity

Green.

The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,

Criterion III, due to all containment cooler drip pans being degraded such that the containment cooler condensate monitoring system could not perform its design basis safety function to quantify reactor coolant system leakage into the containment atmosphere. Wolf Creek initiated Condition Report 24005 and Work Order 10-325741-000 to clean and repair the drip pans.

This issue is more than minor because it was associated with the equipment performance aspect 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. Specifically, it affected the licensees ability to detect a reactor coolant system leak. The inspectors used Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, to analyze the significance of this finding. The inspectors concluded the finding is of very low safety significance because the condition was not related to pressurized thermal shock. The inspectors also determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because Wolf Creek failed to identify adverse postwork conditions after the coolers received maintenance in the 2009 refueling outage

P.1(a) (Section 1R22).

REPORT DETAILS

Summary of Plant Status

The plant started the inspection period at 100 percent rated thermal power. On January 26, Wolf Creek reduced power to 30 percent per Technical Specification 3.0.3 due to missed required snubber inspections. The plant returned to full power on January 27. On March 2, Wolf Creek tripped from 100 percent reactor power due to low-low steam generator level. Wolf Creek restarted on March 8, however, during the subsequent power ascension Wolf Creek tripped from 42 percent reactor power due to a manual trip when main feed Pump A tripped resulting in a loss of all feedwater. Wolf Creek restarted again on March 9 and achieved 100 percent reactor power on March 10. Wolf Creek remained at 100 percent for the remainder of the inspection period.

REACTOR SAFETY

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

1R01 Adverse Weather Protection

.1 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On January 6, 2010, a winter-weather advisory was issued for expected snow squalls.

The inspectors observed the licensees preparations and planning for the significant winter weather potential. The inspectors reviewed licensee procedures and discussed potential compensatory measures with control room personnel. The inspectors focused on plant managements actions for implementing the stations procedures for ensuring adequate personnel for safe plant operation and emergency response would be available. The inspectors conducted a site walkdown including essential service water intake structures and systems to check for maintenance or other apparent deficiencies that could affect system operations during the predicted significant weather which was expected to provide conditions favorable to frazil ice formation. The inspectors also reviewed corrective action program items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one readiness for impending adverse weather condition sample as defined in IP 71111.01-05.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignments

.1 Partial

Walkdown

a. Inspection Scope

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

• January 5, 2010, Turbine-driven auxiliary feedwater during motor-driven auxiliary feedwater Train A maintenance

• January 5, 2010, Motor-driven auxiliary feedwater Train B during motor-driven auxiliary feedwater Train A maintenance

• March 3, 2010, Containment radiation Monitors GTRE-31 and GTRE-32

• March 4-5, 2010, Partial equipment alignment for Emergency Diesel Generator A following restoration from jacket water cooling system maintenance The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Safety Analysis Report (USAR), technical specification requirements, administrative technical specifications, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four partial system walkdown samples as defined in IP 71111.04-05.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

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

• February 10, 2010, 2000 elevation of the control building

• March 2, 2010, 1988 auxiliary feedwater Area A-33

• March 2, 2010, 2026 auxiliary shutdown panel rooms

• March 18, 2010, Lower cable spreading Room 2034

• March 18, 2010, Upper cable spreading Room 2071

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

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants individual plant examination of external events with later additional insights, their potential to affect equipment that could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event. Using the documents listed in the attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed, that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees corrective action program.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of five quarterly fire-protection inspection samples as defined by IP 71111.05-05.

b. Findings

Introduction.

The inspectors identified a Green noncited violation of License Condition 2.C(5)(a) for degraded fire seals that separated redundant safe shutdown equipment.

Description.

On March 10, 2010, while touring Fire Area A33, the inspectors observed degraded fire seals for two penetrations in the 1988 auxiliary feedwater corridor walls.

Specifically, the hotwell to condensate storage tank piping had moved approximately one inch towards the turbine building and damaged the ceramic damming boards and silicone foam. The inspectors could feel air moving into the auxiliary building space from the turbine building. Procedure M-663-00017, Penetration Seal Typical Details, Revision W20, provides design information for installation and inspection of the seals.

Detail drawing and limiting parameters for an M-1 seal specified a pair of one-inch thick

damming boards and nine inches of silicone foam in the required 3-hour rated configuration. The inspectors identified the silicone foam and ceramic fiber board seals separating the auxiliary feedwater trains from the turbine building and the condensate storage tank valve house were degraded so that they no longer provided a 3-hour rated fire barrier. The air moving through the seal indicated separation from the pipe for the entire depth of the seal. No air could be felt moving between the auxiliary feedwater space and the condensate storage tank valve room.

When notified by the inspectors, the licensee initiated Work Requests 10-079244 and 10-079243, Breach Permits 2010-049 and -050, Condition Report 23828, and compensatory actions for the degraded seals. Wolf Creek repaired the two penetrations using Work Orders 10-325727-000 and 10-325733-000. Damming boards were replaced, additional fire insulation was stuffed into the turbine building penetration, and the damaged silicone foam was replaced.

Analysis.

The failure of the fire seal to have been in the required configuration that resulted in degradation of the 3-hour fire barrier between fire areas was a performance deficiency. The finding was more than minor because it was similar to not minor if section of Example 2.e. of NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, as the performance deficiency impacted the ability of the seal to perform its function. In addition, the performance deficiency was associated with the Mitigating Systems Cornerstone attribute of protection against external events, and affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, under Fire Barrier Degradation, Table A2.2, the finding was associated with Moderate B degradation due to the seal not being in a tested or evaluated condition. Using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, in supplemental screening for fire confinement findings, the finding screens as Green due to exposing Fire Area A33 featuring an automatic full area water-based suppression system. No crosscutting aspect was assigned as this condition was not reflective of current licensee performance.

Enforcement.

Wolf Creek License Condition 2.C.(5)(a) requires, in part, that the licensee maintain in effect all provisions of the approved fire protection program. The Wolf Creek fire protection program, as documented by the Updated Safety Analysis Report, Revision 22, Table 9.5A-1, Section D.1.(j), states that where fire barriers are provided to separate redundant safe shutdown trains, piping penetrations are sealed to provide a fire resistance rating of three hours. Contrary to the above, prior to March 10, 2010, wall penetration Seals P125W2319 and P125W2309, which separate redundant safe shutdown trains of motor-driven auxiliary feedwater from the turbine building and the condensate storage tank valve house, were degraded such that they were not sealed to provide a fire resistance rating of three hours. Since the violation was of very low safety significance and was documented in the licensees corrective action program as Condition Report 23828, it is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000482/2010002-01, Degraded Fire Barriers for Auxiliary Feedwater.

1R11 Licensed Operator Requalification Program

a. Inspection Scope

On February 11, 2010, the inspectors observed a crew of licensed operators in the plants simulator to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• Licensed operator performance

• Crews clarity and formality of communications

• Crews ability to take timely actions in the conservative direction

• Crews prioritization, interpretation, and verification of annunciator alarms

• Crews correct use and implementation of abnormal and emergency procedures

• Control board manipulations

• Oversight and direction from supervisors

• Crews ability to identify and implement appropriate technical specification actions and emergency plan actions and notifications The inspectors compared the crews performance in these areas to pre-established operator action expectations and successful critical task completion requirements.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one quarterly licensed-operator requalification program sample as defined in IP 71111.11-05.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

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

• March 10, 2010, Review of condenser offgas radiation Monitor GERE 0092 failures from 2008 to present

• March 19, 2010, Atmospheric relief valves and main steam safety valves The inspectors reviewed events such as where ineffective equipment maintenance has resulted in valid or invalid automatic actuations of engineered safeguards systems and

independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• Implementing appropriate work practices

• Identifying and addressing common cause failures

• Scoping of systems in accordance with 10 CFR 50.65(b)

• Characterizing system reliability issues for performance

• Charging unavailability for performance

• Trending key parameters for condition monitoring

• Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or (a)(2)

• Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as requiring the establishment of appropriate and adequate goals and corrective actions for systems classified as not having adequate performance, as described in 10 CFR 50.65(a)(1)

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two quarterly maintenance effectiveness samples as defined in Inspection Procedure IP 71111.12-05

b. Findings

===.1

=

The inspectors identified a Green noncited violation of 10 CFR 50.65 for failure to establish goals per paragraph (a)(1) to monitor the performance of the main condenser offgas radiation Monitor GERE0092.

Description.

The main condenser offgas radiation Monitor GERE0092 maintenance rule functions are to identify and monitor indications of a primary to secondary leak (Function SP-03) and, if alarmed, to isolate the steam generator blowdown flowpath (Function SP-04). Wolf Creek had established performance criteria for Function SP-04 of less than three functional failures per 18-month rolling cycle.

On May 29, 2009, radiation Monitor GERE0092 could not pull a sample to enable detection of primary to secondary leakage to result in Function SP-04, steam generator blowdown isolation, if necessary. The radiation monitor was restored and then failed again on June 1, 2009, and again on June 2, 2009. Wolf Creek had documented multiple examples of insufficient sample flow caused by water intrusion since December 11, 2008, but had only evaluated the May 29, 2009, event described in Condition Report 17567 as a functional failure. Each of the following occurrences of insufficient sample

flow was an example where the detector would not have been able to respond to primary to secondary leakage:

• December 11, 2008, Condition Report 13797

• February 9, 2009, Condition Report 14609

• May 29, 2009, Condition Report 17524

• May 29, 2009, Condition Report 17567 (three separate failures)

• July 3, 2009, Condition Report 18284

• August 9, 2009, Condition Report 18342

On June 17, 2009, the maintenance rule expert panel meeting minutes stated, in part, that only one SP-04 functional failure had occurred associated with Condition Report 17567 over the period from May 29, 2009, through June 2, 2009. The meeting minutes stated that GERE0092 had a visible amount of water inside the flow gauge which made GERE0092 inoperable due to the water intrusion. This functional failure was similar to the other failures in the examples above, yet only one failure was counted despite not crediting unavailability time between the three failures from May 29, 2009, to June 2, 2009. On March 10, 2010, the inspectors questioned why only one failure in the last 18 months was considered as a functional failure. Also, failures documented in Condition Report 17567 were only counted as unavailability for the SP-04 function and not towards the SP-03 function despite the inability of the radiation monitor to detect primary to secondary leakage. Wolf Creek engineering evaluated the issues identified by the inspectors and determined that Function SP-04 should have been moved to a(1)for goal setting.

Analysis.

Failure to establish monitoring goals in response to unreliable system performance and classify as Maintenance Rule a(1) is a performance deficiency. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events. The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings,"

and determined that this finding is of very low safety significance, Green. Specifically, the associated function (SP-04) to detect primary to secondary leakage and then isolate the steam generator blowdown flow path does not result in a loss of any safety function.

The inspectors determined that this finding has a crosscutting aspect in the problem identification and resolution area associated with corrective action program because Wolf Creek failed to take appropriate corrective actions to address the system reliability issue and adverse radiation monitor performance trends in a timely manner, commensurate with safety significance and complexity P.1(d).

Enforcement.

Title 10 CFR 50.65 (a)(1) requires, in part, that holders of an operating license shall monitor the performance or condition of structures, systems, or components within the scope of the monitoring program against licensee established goals in a manner sufficient to provide reasonable assurance that such structures, systems, or components are capable of fulfilling their intended safety functions. Title 10 CFR 50.65 (a)(2) states, in part, that monitoring as specified in paragraph (a)(1) is not required where it has been demonstrated that the performance or condition of a structure, system, or component is being effectively controlled through performance of appropriate preventive maintenance, such that the structure, system, or component remains capable of performing its intended function. Contrary to the above, between June 17, 2009 and

March 11, 2010, the licensee failed to demonstrate that the performance of safety-related radiation monitor was being effectively controlled through appropriate preventative maintenance. Specifically, after exceeding the safety-related monitor Function SP-04 performance criteria, Wolf Creek failed to evaluate and establish a(1)goals. Because this violation was of very low safety significance and was entered into the licensees corrective action program under Condition Report 24423, this violation is being treated as a noncited violation in accordance with the NRC Enforcement policy:

NCV 05000482/2010002-02, Failure to Establish Goals and Monitor for a(1) Offgas Radiation Monitor GERE0092.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

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

• March 4, 2010, Shutdown risk assessment for Mode 4, emergent work on NI0036

• February 22, 2010, Risk assessment for atmospheric relief valve maintenance during component cooling water heat exchanger testing

• March 2, 2010, Emergent work for Emergency Diesel Generator A jacket water leakage

• March 19, 2010, Emergent work for Emergency Diesel Generator B exhaust silencer degradation The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four maintenance risk assessments and emergent work control inspection samples as defined by IP 71111.13-05.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• January 26, 2010, Degraded snubbers on essential service water system

• March 3, 2010 Terry turbine casing leak

• October 10, 2009, Reactor coolant system overboration

The inspectors selected these potential operability issues based on the risk-significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that technical specification operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the technical specifications and the Updated Safety Analysis Report to the licensees evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations.

Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three operability evaluations inspection samples as defined in IP-71111.15-05

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

.1 Temporary Modifications

a. Inspection Scope

To verify that the safety functions of important safety systems were not degraded, the inspectors reviewed the temporary modification identified as a wiring change to eliminate a potential hot short in the voltage regulator for the Train B emergency diesel generator.

The inspectors reviewed the temporary modification and the associated safety evaluation screening against the system design bases documentation, including the USAR and the technical specifications, and verified that the modification did not adversely affect the system operability/availability. The inspectors also verified that the installation and restoration were consistent with the modification documents and that configuration control was adequate. Additionally, the inspectors verified that the temporary modification was identified on control room drawings, appropriate tags were placed on the affected equipment, and licensee personnel evaluated the combined effects on mitigating systems and the integrity of radiological barriers.

These activities constitute completion of one sample for temporary plant modifications as defined in Inspection Procedure 71111.18-05.

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing

a. Inspection Scope

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

• March 6, 2010, Nuclear Instrument NI-36 replacement

• March 4, 2010, PN09 static switch replacement

The inspectors selected these activities based upon the structure, system, or component's ability to affect risk. The inspectors evaluated these activities for the following (as applicable):

• The effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed

• Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate The inspectors evaluated the activities against the technical specifications, the Updated Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with postmaintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two postmaintenance testing inspection sample(s) as defined in IP 71111.19-05.

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors reviewed the outage safety plan and contingency plans for the forced outages conducted from March 2-8 and 8-9, 2010, to confirm that licensee personnel

had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense in depth

• Configuration management, including maintenance of defense in depth, is commensurate with the outage safety plan for key safety functions and compliance with the applicable technical specifications when taking equipment out of service.

• Clearance activities, including confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing.

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

• Status and configuration of electrical systems to ensure that technical specifications and outage safety-plan requirements were met, and controls over switchyard activities.

• Monitoring of decay heat removal processes, systems, and components.

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

• Controls over activities that could affect reactivity.

• Startup and ascension to full power operation, tracking of startup prerequisites, walkdown of the containment to verify that debris had not been left which could block emergency core cooling system suction strainers.

• Licensee identification and resolution of problems related to forced outage activities.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two forced outage inspection samples as defined in IP 71111.20-05.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the Updated Safety Analysis Report, procedure requirements, and technical specifications to ensure that the three surveillance activities listed below demonstrated that the systems, structures, and/or components tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the significant surveillance test attributes were adequate to address the following:

• Preconditioning

• Evaluation of testing impact on the plant

• Acceptance criteria

• Test equipment

• Procedures

• Jumper/lifted lead controls

• Test data

• Testing frequency and method demonstrated technical specification operability

• Test equipment removal

• Restoration of plant systems

• Fulfillment of ASME Code requirements

• Updating of performance indicator data

• Engineering evaluations, root causes, and bases for returning tested systems, structures, and components not meeting the test acceptance criteria were correct

• Reference setting data

• Annunciators and alarms setpoints The inspectors also verified that licensee personnel identified and implemented any needed corrective actions associated with the surveillance testing.

• February 11, 2010, STS Manual Auto Fast Start, Sync & Loading of EDG NE02

• February 21, 2010, STS AB-201, Atmospheric Relief Valve Stroke Testing

• March 3, 2010, Containment cooler condensate collection system for reactor coolant system leakage monitoring Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three surveillance testing inspection samples as defined in IP 71111.22-05.

b. Findings

Introduction.

The inspectors identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, due to all containment cooler drip pans being degraded such that condensate could not be properly collected to meet the requirements of the design basis.

Description.

On March 3, 2010, the inspectors walked down the containment coolers.

The inspectors found that all four coolers had degraded condensate collection drip pans at the ends of the tube bundles. The pans had accumulated appreciable rust and debris, and the sheet metal was bent such that the condensate was running out of the pans instead of flowing to the standpipe for reactor coolant system leakage monitoring. Some of the pans consisted of multiple sections of sheet metal which were not jointed together such that condensate could be properly collected. Once identified by the inspectors, Wolf Creek operators appropriately entered Technical Specification 3.4.15 and performed Surveillance Requirement 3.4.15.1 per Required Action C.1. Surveillance Requirement 3.4.15.1 requires a channel check of the containment atmosphere particulate radiation monitors. Wolf Creek initiated Condition Report 24005 and Work Order 10-325741-000 to clean and repair the drip pans. Containment coolers were inspected during the previous outage, which the inspectors concluded was a missed opportunity to identify and correct this condition.

During the walkdowns, the inspectors noted surface corrosion on ventilation ducts, cable trays, and other equipment located under the containment coolers. The surface corrosion was extensive enough to show that the lack of collection had existed for some time. From the accumulation of rust debris and fiberglass insulation foreign material in the collection pans, the inspectors concluded the condition of the pans went unnoticed during cooler maintenance performed during the fall 2009 refueling outage. Upon reviewing the design basis drawings the inspectors determined that the as-found condition of the drip pans did not meet the design basis.

The inspectors concluded that Technical Specification 3.4.15.1, Condition C.1, was met because control room operators perform Surveillance Requirement 3.4.15.1 three times every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The inspectors interviewed operators and reviewed a sample of completed control room surveillance Procedure STS CR-001 and determined that operators regularly implement the surveillance procedure. This surveillance procedure satisfies Surveillance Requirement 3.4.15.1. The inspectors also found past instances in which both containment particulate radiation monitors were inoperable simultaneously with the condensate monitoring system placing Wolf Creek in Condition D.1, but the radiation monitors were inoperable for much less than the 30-day completion time. The inspectors concluded that Wolf Creek complied with Conditions C.1 and D.1 while the containment air cooler condensate monitoring system was inoperable.

Analysis.

The failure to maintain the drip pans in accordance with plant design basis station drawings is a performance deficiency. This issue is more than minor because it was associated with the equipment performance aspect 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. Specifically, it affected the licensees ability to detect a reactor coolant system leak. The inspectors used Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, to analyze the significance of this finding. The inspectors concluded the finding is of very low safety significance because the condition was not related to pressurized thermal shock. The inspectors also determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because Wolf Creek failed to identify adverse postwork conditions after the coolers received maintenance in the 2009 refueling outage P.1(a).

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that design control measures shall provide for verifying or checking the adequacy of design. Wolf Creek monitors reactor coolant system pressure boundary leakage to containment atmosphere, in part, through the containment cooler condensate monitoring system. Contrary to the above, prior to March 3, 2010, the containment cooler condensate monitoring system no longer met the established design basis because the drip pans were degraded such that they could not correctly collect condensate for reactor coolant system leakage monitoring. Because this issue was determined to be of very low safety significance (Green) and was entered into the licensees corrective action program as Condition Report 23885, this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy:

NCV 05000482/2010002-03, Containment Cooler Condensate Monitoring System Not In Accordance with Design Basis Station Drawings.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspectors performed an in-office review of the Wolf Creek Radiological Emergency Response Plan, Revision 9, and the Emergency Action Level Procedure APF 06-002-01, Revision 13. The revision to the emergency plan included facility and position name changes, changes to the method of transmitting news releases, and updating sampling location maps. The revision to the emergency action level form included adding the containment purge isolation system as a containment isolation closure function check and updating the emergency action levels for modifications performed on the plants seismic monitoring system.

The revisions were compared to previous revisions, 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, to NUMARC/NESP-007, Methodology for Development of Emergency Action Levels, Revision 2, 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). 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 two samples as defined in Inspection Procedure 71114.04-05.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of a routine licensee emergency drill on March 23, 2010, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed

emergency response operations in the simulator to determine whether the event classification, notifications, and protective action recommendations were performed in accordance with procedures. The inspectors also attended the licensee drill critique to compare any inspector-observed weakness with those identified by the licensee staff in order to evaluate the critique and to verify whether the licensee staff was properly identifying weaknesses and entering them into the corrective action program. As part of the inspection, the inspectors reviewed the drill package and other documents listed in the attachment.

These activities constitute completion of one sample as defined in IP 71114.06-05.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the data submitted by the licensee for the 4th Quarter 2009 performance indicators for any obvious inconsistencies prior to its public release in accordance with Inspection Manual Chapter 0608, Performance Indicator Program.

This review was performed as part of the inspectors normal plant status activities and, as such, did not constitute a separate inspection sample.

b. Findings

No findings of significance were identified.

.2 Unplanned Scrams per 7000 Critical Hours (IE01)

a. Inspection Scope

The inspectors sampled licensee submittals for the unplanned scrams per 7000 critical hours performance indicator for the period from the 1st quarter 2009 through the 4th quarter 2009. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5.

The inspectors reviewed the licensees operator narrative logs, issue reports, event reports, and NRC Integrated Inspection Reports for the period of January 1, 2009, through December 31, 2009, to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one unplanned scrams per 7000 critical hours sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.3 Unplanned Scrams with Complications

a. Inspection Scope

The inspectors sampled licensee submittals for the unplanned scrams with complications performance indicator for the period from the 1st quarter 2009 through the 4th quarter 2009. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees operator narrative logs, issue reports, event reports, and NRC Integrated Inspection Reports for the period of January 1, 2009, through December 31, 2009, to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report. Wolf Creek has submitted a frequently asked question on this indicator.

These activities constitute completion of one unplanned scrams with complications sample as defined in IP 71151-05.

b. Findings

No findings of significance were identified

.4 Unplanned Power Changes per 7000 Critical Hours

a. Inspection Scope

The inspectors sampled licensee submittals for the unplanned power changes per 7000 critical hours performance indicator for the period from the 1st quarter 2009 through the 4th quarter 2009. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees operator narrative logs, issue reports, maintenance rule records, event reports and NRC Integrated Inspection Reports for the period of January 1, 2009, through December 31, 2009, to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one unplanned power changes per 7000 critical hours sample as defined in IP 71151-05.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

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

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees corrective action program at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. The inspectors reviewed attributes that included: the complete and accurate identification of the problem; the timely correction, commensurate with the safety significance; the evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrences reviews; and the classification, prioritization, focus, and timeliness of corrective actions. Minor issues entered into the licensees corrective action program because of the inspectors observations are included in the attached list of documents reviewed.

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

b. Findings

No findings of significance were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow up, the inspectors performed a daily screening of items entered into the licensees corrective action program. The inspectors accomplished this through review of the stations daily corrective action documents.

The inspectors performed these daily reviews as part of their daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings of significance were identified.

.3 Selected Issue Follow Up Inspection

a. Inspection Scope

During a review of items entered in the licensees corrective action program, the inspectors recognized a corrective action item documenting inspectors review of Wolf Creeks disposition of a parts discrepancy regarding the packing gland follower on Valve EG-HV-102.

These activities constitute completion of one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

No findings of significance were identified.

4OA3 Event Follow-up

.1 Reactor Downpower Due to Missed Technical Requirements Manual Action Statement

a. Inspection Scope

On January 26, 2010, Wolf Creek reduced power to 30 percent after entering Technical Specification 3.0.3. This was due to failure to meet Technical Requirements Manual Technical Requirement 3.7.20, Required Action C.1. Action C.1 required inspection of snubbers within six months of a system transient. At the time of discovery, this action statement had not been implemented for any transient in the essential service water system. With Action Statement C.1 not met, Action Statement D.1 required operators to immediately declare the essential service water system inoperable, which caused the operators to enter Technical Specification 3.0.3. Wolf Creek made a notification to the headquarters operations officer for a technical specification required shutdown per 10 CFR 50.72(b)(2), and a safety system functional failure for loss of equipment needed to shutdown the reactor and mitigate accidents per 10 CFR 50.72(b)(3)(v)(A) and (D). The inspectors reviewed the 10 CFR 50.72 notifications and observed the power reduction in the control room and observed no problems.

Mechanics inspected a total of 19 snubbers in the containment and the auxiliary building.

Train B essential service water had a higher priority because it had fewer affected snubbers. At 8:13 p.m., Wolf Creek declared all essential service water system Train B snubbers satisfactory and exited Technical Specification 3.0.3. Wolf Creek continued in Technical Specifications 3.7.8 and 3.8.1 for Train A essential service water inoperable.

Train A essential service water snubbers were later found satisfactory as well. Wolf Creek then began to increase reactor power.

These activities constitute completion of one event followup sample as defined in IP 71153-05.

b. Findings

No findings of significance were identified.

.2 Reactor Trip on Lo-Lo Steam Generator Level After Loss of Nonvital Instrument

Bus PN09

a. Inspection Scope

The inspectors reviewed the equipment performance and operator response associated with an automatic reactor trip from full power on March 2, 2010, caused by low steam generator levels. Prior to the reactor trip, maintenance was planning on performing a light bulb change. Inverter PN09 was successfully swapped to the alternate source and the light bulb changed on the normal power source. After realigning to the normal power source, voltage oscillations started approximately 2 seconds later and a fuse failed causing loss of power to the bus. The control room received several annunciators powered by PN09. The balance of plant operator noted 0 rpm on main feedwater Pump A, took manual control and tried to raise pump speed. Main feedwater Pump A tripped approximately 8 seconds after loss of PN09. Steam generator levels began to decrease. Operators reset main feedwater Pump A from its tripped condition. The shift manager ordered a manual reactor trip, but the reactor automatically tripped on lo-lo steam generator level prior to the operator turning the reactor trip handswitch 69 seconds after the failure of PN09.

These activities constitute completion of one event followup sample as defined in IP 71153-05.

1. Failure to Perform Adequate Post Trip Review

Introduction.

The inspectors identified a Green finding for the failure to adequately implement the posttrip review procedure. Specifically, Wolf Creek failed to adequately evaluate the events leading to the reactor trip that occurred on March 2, 2010.

Description.

On March 2, 2010, Wolf Creek Generating Station experienced an automatic reactor trip from full power. The trip initiated from lo-lo steam generator water levels following the trip of main feedwater Pump A. The plant transient began due to a loss of nonsafety-related Inverter PN09 which supplies power to the main feedwater Pump A speed control circuitry. The controller also receives 120Vac power from another source which did not fail. Following a reactor trip, Wolf Creek is required to perform a posttrip review evaluation in accordance with Procedure AP 20-002, Post-Trip Review, Revision 8. As specified in Generic Letter 83-28, Required Actions Based on Generic Implications of Salem ATWS Events, this procedure requires a thorough and systematic evaluation of reactor trip events to ensure that the causes, as well as the responses of safety-related equipment, are fully understood prior to plant restart.

The inspectors reviewed Wolf Creek's posttrip review and determined that the evaluation was not adequate because it failed to identify that some aspects of operator performance influenced the automatic response of a mitigating system during the transient. The inspectors determined that the licensees conclusion that the loss of Inverter PN09 directly resulted in a trip of main feedwater Pump A was not consistent with the actual pump performance. Data traces confirmed that following the loss of PN09, main feedwater Pump A did not immediately trip, but instead slowed then accelerated before tripping on overspeed several seconds later. The licensees conclusion was also inconsistent with operator statements that indicated the balance of

plant reactor operator took manual control of main feedwater Pump A following the loss of PN09. The inspectors determined that because the loss of PN09 resulted in a loss of speed indication for the main feedwater Pump A, the operator incorrectly assumed the controller had failed and attempted to increase speed manually; however, it is not clear whether this occurred before or after the pump had tripped on overspeed. Plant computer data verified a large increase in feedwater pump suction flow (and indirectly speed) occurred prior to the pump trip. Also, the plant computer alarm for low suction pressure on main feedwater pump A was received prior to pump trip.

Additionally, the inspectors determined that the Wolf Creeks posttrip review failed to identify that operator response to the loss of main feedwater Pump A was not in accordance with station procedures. During a review of the posttrip data, the inspectors noted that control room operators reset main feedwater Pump A prior to the reactor trip.

Actions should have been taken by the licensed operators per Procedure ALR 00-120A, MFP A Trip, Step 3.a, which directs the operators to decrease reactor power to less than or equal to 62 percent per Procedure OFN MA-038, Rapid Plant Shutdown.

Wolf Creek evaluated the individual issues and deficiencies listed above and entered them into the corrective action program as Condition Reports 23932, 23966, 24043, 23982, and 23981.

Analysis.

The performance deficiency associated with this finding involved a failure to adequately follow procedures for conducting posttrip reviews. This finding was greater than minor because the information omitted from the posttrip review was associated with the human performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since the finding does not represent a loss of system safety function, nor does the finding represent actual loss of safety function for single train for a greater time than permitted by technical specifications. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because Wolf Creek failed to fully evaluate plant computer data and operator statements associated with the March 2, 2010, reactor trip P.1(c).

Enforcement.

Enforcement action does not apply because the performance deficiency did not involve a violation of regulatory requirements. This finding is of very low safety significance and the issue was entered into the licensee's corrective action program as Condition Reports 23932, 23966, 24043, 23982, and 23981:

FIN 05000482/2010002-04, Failure to Perform Adequate Post Trip Review.

2. Failure to Follow Procedure for a Main Feed Pump Trip

Introduction.

The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a, Procedures, for the failure of Wolf Creek control room personnel to follow procedures for a main feedwater pump trip.

Description.

On March 2, 2010, Wolf Creek maintenance performed work on nonsafety-related Inverter PN09. Due to a faulty static switch control board, a total loss of inverters 120Vac instrument voltage occurred which resulted in an apparent loss of the main feedwater Pump A speed control circuitry. The pump tachometer read zero rpm. The operator took manual control of the pump and attempted to increase

speed; however, it is unclear whether or not the pump had already tripped on overspeed. Plant computer data shows the pump suction flow decreased and then rapidly increased, then the pump tripped. During this transient that followed, the operator also reset the main feedwater Pump A trip eight seconds prior to the reactor trip.

Procedural reviews revealed that these actions were not performed in accordance with the station procedures for a loss of one main feed pump. Action should have been taken by the licensed operators per Procedure ALR 00-120A, MFP A Trip, Step 3.a, which directs the operators to decrease reactor power to less than or equal to 62 percent per Procedure OFN MA-038, Rapid Plant Shutdown. Subsequent Wolf Creek simulator evaluations indicate that the trip would not have been prevented once PN09 was lost. The loss of PN09 compromised the condenser available interlock, rendering the steam dumps unavailable. Therefore, a load reduction would have compressed the steam bubbles in the steam generators, causing the generators to reach the lo-lo trip setpoint level. Thus, had the operators followed the rapid plant shutdown procedure, the plant would have tripped sooner. In this case, however, the tripped main feedwater pump would not have been reset and the auxiliary feedwater system would be capable of receiving an auto start actuation on a trip of main feedwater Pump B as described below. This issue was entered into the Wolf Creek corrective action program as Condition Report 24011.

The engineered safety feature actuation system logic relies on the main feedwater pump control oil pressure switches to actuate auxiliary feedwater. Resetting of the main feedwater pumps incorrectly provided an input to the logic that that pump is supplying water to the steam generators. Wolf Creek Technical Specification 3.3.2 Engineered Safety Feature Actuation System Instrumentation, Table 3.3.2-1, Function 6.G, requires two main feedwater trip channels per pump to be in operation while in Modes 1 and 2.

Technical Specification Limiting Condition Operation 3.3.2, Condition J, allowed for one main feedwater pump trip channel to be out of service, but no provision is provided for multiple channels being out of service. The action taken by the reactor operator resulted in both channels of main feedwater Pump A trip circuitry being out of service for both trains of auxiliary feedwater actuation system Function g. In this condition (one pump reset but not providing water to the steam generators) the auxiliary feedwater system would be prevented from receiving an auto start actuation on a loss of all feedwater caused by a trip main feedwater Pump B. Since Limiting Condition for Operation 3.3.2, Condition J, was not met and an associated action was not provided, the inspectors determined that Technical Specification Limiting Condition for Operation 3.0.3 should have applied.

Wolf Creek had previously identified the potential to impact the auxiliary feedwater actuation circuitry due to the reset of a tripped main feedwater pump. Condition Report 23008 was issued January 25, 2010, to document operating experience that identified that the design and operation of the main feedwater pump instrumentation may not be consistent with the technical specification requirements. The evaluation was completed and essential required reading was developed and issued on February 18, 2010, for all licensed reactor operators and senior reactor operators.

Analysis.

Control room personnel failing to follow station procedures in the event of a main feedwater pump trip is a performance deficiency. This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of

human performance and it affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and screened the finding to Phase 2 because the finding represents a loss of auxiliary feedwater actuation system safety Function g. The finding screened to Phase 3 because of the failure to start of both motor-driven auxiliary feedwater pumps. The senior reactor analyst performed a Phase 3 analysis and concluded that the finding was Green because the probability of an initiator occurring within any 10-second exposure time is approximately 3E-7. Additionally, auxiliary feedwater pumps would have been automatically started on lo-lo steam generator level if required. The inspectors also determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with operating experience because Wolf Creek failed to communicate relevant operating experience to affected internal stakeholders P.2(a).

Enforcement.

Technical Specification 5.4.1.a, Procedures, requires that written procedures be established and implemented covering activities specified in Appendix A of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation),

February 1978. Regulatory Guide 1.33, Appendix A, Section 6.j, required operating procedures for loss of feedwater or feedwater system failures. Step 3.a of Procedure ALR 00-120A, MFP A Trip, Revision 8, required the operator to decrease reactor power to less than or equal to 62 percent per Procedure OFN MA-038, Rapid Plant Shutdown. Contrary to the above, on March 2, 2010, operators did not decrease reactor power to less than or equal to 62 percent per Procedure OFN MA-038 following the trip of main feedwater pump A. Specifically, the reactor operator took action to manually control and reset the main feedwater Pump A. Because this finding is of very low safety significance and the licensee has entered this issue into the corrective action program as Condition Report 24011, this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy:

NCV 05000482/2010002-05, Failure to Follow Procedure for a Main Feed Pump Trip.

.3 Manual Reactor Trip after Trip of Main Feed Pump A

a. Inspection Scope

The inspectors reviewed the equipment performance and operator actions associated with a manual reactor trip from 43 percent power when main feedwater Pump A tripped during plant startup on March 8, 2010. Main feedwater Pump A had been in service since low power when the motor-driven feed pump was stopped. Wolf Creek later determined that main feedwater Pump A tripped on overspeed when a defective controller card drove the speed signal high. The speed was sustained because the feed pump turbine pilot valve stuck open.

These activities constitute completion of one event followup sample as defined in IP 71153-05.

b. Findings

No findings of significance were identified.

.4 (Closed.) Licensee Event Report 2008-001-00, Completion of a Technical Specification

Shutdown.

On January 10, 2008, Wolf Creek reduced power and entered Mode 3 due to noncondensable gas accumulation in the suction of the safety injection pump suction and discharge lines. NRC Region IV sent a special inspection team on January 16, 2008, to review this issue. The inspection results are documented in NRC Inspection Report 05000482/2008006. This LER is closed.

4OA6 Meetings

Exit Meeting Summary

On February 4, 2010, the emergency preparedness inspector conducted a telephonic exit meeting to present the results of the in-office inspection of changes to the licensees emergency plan and emergency action levels to Mr. T. East. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On April 8, 2010, the resident inspectors presented the inspection results of the inspections to Mr. M. Sunseri, President and Chief Executive Officer, and other members of the licensee's management staff. The licensee acknowledged the findings presented. The inspectors noted that while proprietary information was reviewed, none was retained or included in this report.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

M. Sunseri, President and Chief Executive Officer

S. E. Hedges, Site Vice President

R. Gardner, Plant Manager

T. East, Superintendent, Emergency Preparedness

P. Bedgood, Superintendent, Chemistry/Radiation Protection

G. Pendergrass, Manager, System Engineering

L. Ratzlaff, Supervisor, Support Engineering

G. Neises, Manager, Design Engineering

S. Koenig, Manager, Corrective Action

S. Henry, Manager, Operations

B. Dale, Manager, Maintenance

D. Dees, Supervisor, Operations Support

R. Flannigan, Manager, Regulatory Affairs

D. Hooper, Supervisor, Licensing

J. Simmons, Maintenance Rule Engineer

T. Slenker, Operations Support Engineer

S. Wideman, Senior Licensing Engineer

S. Atkin, Design Engineer

L. Rockers, Licensing Engineer

B. Muilenberg, Licensing Engineer

NRC Personnel

B. Singal, Senior Project Manager, NRR

R. Deese. Senior Project Engineer, Region IV

G. Tutak, Project Engineer, Region IV

LIST OF ITEMS

OPENED AND CLOSED

Opened and Closed

05000482/2010002-01 NCV Degraded Fire Barriers for Auxiliary Feedwater (Section 1R05)

05000482/2010002-02 NCV Failure to Establish Goals and Monitor for a(1) Offgas Radiation Monitor GERE0092 (Section 1R12)

05000482/2010002-03 NCV Inoperable Containment Cooler Condensate Monitoring System (Section 1R22)

Opened and Closed

05000482/2010002-04 FIN Failure to Perform Adequate Post Trip Review (Section 4OA3.2)

05000482/2010002-05 NCV Failure to Follow Procedure for a Main Feed Pump Trip (Section 4OA3.2)

Closed

2008-001-00 LER Completion of a Technical Specification Required Shutdown (Section 4OA3.6)