IR 05000482/2008004

IR 05000482-08-004; on 6/29/2008 - 9/27/2008; Wolf Creek; Equipment Alignment, Maintenance Risk and Emergent Work, Operability Evaluations, Postmaintenance Testing, Problem Identification and Resolution, Event Followup, and Other Activities

ML083120336
Person / Time
Site:	Wolf Creek
Issue date:	11/07/2008
From:	Vincent Gaddy NRC/RGN-IV/DRP/RPB-B
To:	Muench R A Wolf Creek
References
IR-08-004
Download: ML083120336 (43)

Text

November 7, 2008

Rick A. Muench, President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839

SUBJECT: WOLF CREEK GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000482/2008004

Dear Mr. Muench:

On September 27, 2008, the U.S. Nuclear Regulatory Commission completed an integrated inspection at your Wolf Creek Generating Station. The enclosed report documents the inspection results, which were discussed on October 3 and 15, 2008, with Mr. M. Sunseri and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of this inspection, four NRC-identified and four self-revealing findings of very low safety significance (Green) are documented in this report. All of these findings were determined to involve violations of NRC requirements. However, because of the very low safety significance and because the findings were entered into your corrective action program, the NRC is treating these violations as noncited violations consistent with Section VI.A of the NRC Enforcement Policy. If you contest these noncited violations, you should provide a response within 30 days 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. Wolf Creek Nuclear Operating Corp. - 2 -

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/

Vincent G. Gaddy, Chief Project Branch B Division of Reactor Projects Docket No. 50-482 License No. NPF-42

Enclosure:

Inspection Report 05000482/2008004

w/Attachment:

Supplemental Information cc w/enclosure: Vice President Operations/Plant Manager Wolf Creek Nuclear Operating Corp.

P.O. Box 411 Burlington, KS 66839 Jay Silberg, Esq. Pillsbury Winthrop Shaw Pittman LLP 2300 N Street, NW Washington, DC 20037 Supervisor Licensing Wolf Creek Nuclear Operating Corp. P.O. Box 411 Burlington, KS 66839 Chief Engineer, Utilities Division Kansas Corporation Commission 1500 SW Arrowhead Road Topeka, KS 66604-4027 Office of the Governor State of Kansas Topeka, KS 66612 Attorney General 120 S.W. 10th Avenue, 2nd Floor Topeka, KS 66612-1597 County Clerk Coffey County Courthouse 110 South 6th Street Burlington, KS 66839-1798 Chief, Radiation and Asbestos Control Section Kansas Department of Health and Environment Bureau of Air and Radiation 1000 SW Jackson, Suite 310 Topeka, KS 66612-1366 Ronald L. McCabe, Chief Technological Hazards Branch National Preparedness Division DHS/FEMA 9221 Ward Parkway Suite 300 Kansas City, MO 64114-3372 Wolf Creek Nuclear Operating Corp. - 3 -

Electronic distribution by RIV: Regional Administrator (Elmo.Collins@nrc.gov) DRP Director (Dwight.Chamberlain@nrc.gov) DRP Deputy Director (Anton.Vegel@nrc.gov) DRS Director (Roy.Caniano@nrc.gov) DRS Deputy Director (Troy.Pruett@nrc.gov)

Senior Resident Inspector (Chris.Long@nrc.gov) SRI, Callaway (David.Dumbacher@nrc.gov) Branch Chief, DRP/B (Vincent.Gaddy@nrc.gov) Senior Project Engineer, DRP/B (Rick Deese@nrc.gov) Acting Team Leader, DRP/TSS (Douglas.Bollock@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) DRS STA (Dale.Powers@nrc.gov) Senior Enforcement Specialist (Mark.Haire@nrc.gov) OEDO RIV Coordinator (Mark.Cox@nrc.gov) R. Kahler, NSIR/DPR/EDP (Robert.Kahler@nrc.gov) ROPreports Resource OEMail Resource WC Site Secretary (Shirley.Allen@nrc.gov)

SUNSI Review Completed: ___VGG__ ADAMS: Yes No Initials: __VGG__

Publicly Available Non-Publicly Available Sensitive Non-Sensitive R:\_REACTORS\_WC\2008\WC 2008-004RP-CML-vgg.doc ML 083120336 SRI:DRP/B SPE:DRP/B C:DRS/EB1 C:DRS/EB2 CMLong RDeese RBywater NOkeefe

/RA/ VGG for /RA/ /RA/ KDC for /RA/ 10/ /2008 10/14/2008 10/24/2008 10/24/2008 C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C:DRP/B RELantz MPShannon GWernerVGGadd y /RA/ /RA/ /RA/ /RA/ 10/24/2008 10/24/2008 10/24/2008 11/17/2008 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 50-482 License: NPF-42 Report: 5000482/2008004 Licensee: Wolf Creek Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane SE Burlington, Kansas Dates: June 29 to September 27, 2008 Inspectors: S. Cochrum, Senior Resident Inspector C. Long, Senior Resident Inspector D. Dumbacher, Callaway Senior Resident Inspector P. Alter, Senior Training Program Specialist R. Hickok, Reactor Technology Instructor J. Groom, Callaway Resident inspector W. Schaup, Acting Resident Inspector R. Deese, Senior Project Engineer

Approved by: V. Gaddy, Chief, Project Branch B Enclosure

SUMMARY OF FINDINGS

IR 05000482/2008004; 6/29/08 - 9/27/08; Wolf Creek Generating Station; Equipment Alignment, Maintenance Risk and Emergent Work, Operability Evaluations, Postmaintenance Testing, Problem Identification and Resolution, Event Followup, and Other Activities.

This report covered a 3-month period of inspection by resident inspectors and regional specialists. The inspection identified eight Green findings, all of which are noncited violations. 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's 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: Initiating Events

Green.

A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified due to an inadequate vent path for the reactor vessel head. The inadequate vent path resulted in the formation of a maximum void size of 2600 gallons in the reactor vessel head on March 23, 2008, while the plant was shutdown and depressurized. Wolf Creek found indirect evidence of a loop seal due to water that came out of the hard pipe portion of the vent rig at the end of the outage during vacuum filling of the reactor coolant system. However, the licensee could not exclude blockage in the piping. This issue was entered into the corrective action program and the licensee plans to conduct a more thorough inspection of the piping during the next refueling outage. The inspectors determined that the failure to provide adequate vessel head vent path to prevent gas accumulation in the reactor vessel during depressurized plant operations was a performance deficiency. The inspectors determined that this finding, which was associated with the initiating events cornerstone, was more than minor because if left uncorrected, it would have become a more significant safety concern. Specifically, without an adequate vent path the reactor vessel does not have an effective means of relieving noncondensable gases to prevent a loss of reactor coolant system inventory. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix G, Attachment 1, and determined it be of very low safety significance based upon the demonstrated availability of mitigating systems and the flooded reactor cavity inventory (Section 4OA3).

Cornerstone: Mitigating Systems

Green.

A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," occurred on April 7, 2008, when a loss of offsite power caused the service water pumps to shut down and the essential service water pump to start. As a result, a water hammer occurred and the control room air conditioning unit Condenser B developed an approximately 80 gpm essential service water leak due to inadequate yield strength of the studs. This issue was entered into the corrective action program as Condition Report 2008-001450. Wolf Creek's operation of the control room air conditioning and essential service water systems outside the design limits of the heat exchanger studs was determined to be a performance deficiency. The finding was determined to be more than minor because it impacted the mitigating systems cornerstone objective of ensuring the availability, reliability, and operability 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 because it did not cause the loss of safety function and did not impact risk for external events. The inspectors determined that the finding has a problem identification and resolution crosscutting aspect in the area associated with the corrective action program. Specifically, Wolf Creek previously identified that the heat exchanger joint might be inadequate, but failed to perform any subsequent corrective action P.1.d] (Section 4OA2).

Green.

The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to account for the effect of emergency diesel generator frequency variation at the lower limit of the allowable range. Specifically, emergency diesel generator voltage and frequency deviations for load sequencing was based on nominal 60 hertz operation of pumps and fans and did not account for the two percent variation allowed by Technical Specification 3.8.1. Wolf Creek could not demonstrate compliance with Updated Safety Analysis Report, Section 8.1.4.3.b. The licensee has entered this issue into their corrective action program as Condition Report 2008-004312. The inspectors determined that failure to properly account for the effect of frequency variation on the emergency diesel generator was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to account for the frequency variations at the lower limit had more than a minimal effect on the outcome of the analysis, in that, the bus frequency will decrease below the Updated Safety Analysis Report limit of 57 hertz for loss of coolant accident and loss of offsite power scenarios. This finding screened as having very low safety significance because it was a design or qualification deficiency confirmed not to result in loss of operability (Section 4OA5).

Green.

The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," because Wolf Creek failed to adequately demonstrate that submerged 4160 volt cables are qualified for such service, and that they will remain operable, although the cables are presently operable. Since NRC Information Notice 2002-12 was issued, Wolf Creek had several opportunities to implement a preventive maintenance program and/or thoroughly evaluate the submerged cables. These cables include those of residual heat removal, containment spray, and essential service water.

Wolf Creek has subsequently written Condition Report 2008-5073 and work orders to inspect cables and dewater cable vaults. Failure to perform an adequate engineering evaluation that demonstrated continued operability was considered a performance deficiency. The inspectors determined that this finding was more than minor using Inspection Manual Chapter 0612, Appendix E, Example 3.j, because the NRC was able to show that Wolf Creek's operability evaluation needed significant change to demonstrate continued operability. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance. Specifically, the deficiency did not result in the present loss of operability or functionality and did not represent a risk significant external event such as flooding. The inspectors determined that the finding has a problem identification and resolution crosscutting aspect in the area associated with the corrective action program. Despite several opportunities since 2002, Wolf Creek failed to perform a thorough evaluation for continued operability of submerged safety-related cables to ensure continued nuclear safety P.1(c) Section 1R15).

Green.

Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," when Wolf Creek mechanically agitated the side of check Valve EP8818D such that the body of the valve was dented numerous times. This activity was performed under a troubleshooting work order to reduce valve seat leakage. The subsequent evaluation stated that this was an acceptable practice and that it would strengthen the surface metal of the valve body. Wolf Creek subsequently initiated Condition Report 2008-2284 to evaluate the practice. The inspectors determined that the failure to utilize work instructions appropriate to the circumstances and properly evaluate the effects was a performance deficiency. The inspectors determined that this issue is more than minor because it could become a more safety-significant concern if the cold working or peening practice is continued. Inspectors determined that the finding was not appropriate for evaluation under Inspection Manual Chapter 0609, Attachment 4. The inspectors applied Inspection Manual Chapter 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria." The inspectors used a bounding qualitative case, and in consultation with NRC management, determined that the operability of the valve was not impacted. Therefore, the finding was determined to be of very low safety significance. The inspectors determined that the finding has a problem identification and resolution crosscutting aspect in the area associated with the corrective action program because the licensee failed to evaluate the problem of seat leakage such that the resolution (a hammer) appropriately addressed the possible causes of valve seat leakage P.1(c) (Section 1R13).

Green.

On April 7, 2008, the inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," due to an approximately 10-15 gpm leak on the Emergency Diesel Generator B lubricating oil heat exchanger cover plate. Water hammer caused part of the cover plate gasket to be ejected from the heat exchanger and created the leak. The inspectors found that the work order to assemble the heat exchanger was inadequate. Wolf Creek evaluations did not identify that vendor manual steps were not incorporated into the installation work order which led to loose cover plate nuts which caused the leak. Wolf Creek subsequently wrote Condition Report 2008-004982. Wolf Creek's failure to ensure that the configuration of both emergency diesel generator lube oil heat exchangers was per plant design was considered a performance deficiency.

Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance. Specifically, the deficiency did not result in the loss of operability or functionality and did not represent a risk significant external event such as flooding. The inspectors determined that the finding has a human performance crosscutting aspect in the area associated with resources. Specifically, Wolf Creek did not ensure that Work Order 08-305289-000 was adequate to ensure nuclear safety by including vendor instructions or acceptance criteria for both emergency diesel generator lube oil heat exchanger cover plates H.2(c) (Section 1R19).

Cornerstone: Barrier Integrity

Green.

A self-revealing noncited violation of Technical Specification 5.4.1.a was identified for the failure to close Valve EC-V025 during a lineup to recirculate the refueling water storage tank through the spent fuel pool cleanup system. These two systems were cross-connected for approximately 5 minutes on July 26, 2008, which resulted in approximately 1500 gallons of spent fuel pool water being inadvertently transferred to the refueling water storage tank. The licensee entered this issue into their corrective action program as Condition Report 2008-003663. Failure to completely close Valve EC-V025 was a performance deficiency. This finding is more than minor because it is associated with the barrier integrity cornerstone attribute of configuration control and affected the cornerstone objective to maintain functionality of the spent fuel pool system. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance because the finding affected only the barrier function of the spent fuel pool. The inspectors determined that the finding has a problem identification and resolution crosscutting aspect in the area associated with the corrective action program because Wolf Creek did not take appropriate corrective actions to address the adverse trend in manual valve stem friction in a timely manner, commensurate with its safety significance and complexity P.1(d) (Section 40A2).

Green.

A self-revealing noncited violation of 10 CFR Part 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified after the licensee followed two incompatible procedures simultaneously resulting in the inadvertent partial draining of the spent fuel pool. Consequently, approximately 6400 gallons of water was pumped from the spent fuel pool to the refueling water storage tank. Wolf Creek subsequently initiated Condition Report 2008-002035.

Failure to prevent spent fuel pool draining due to simultaneous performance of incompatible Procedures SYS EC-200 and SYS EC-320 is considered a performance deficiency. This finding is more than minor because it impacted the barrier integrity cornerstone attribute of configuration control and affected the cornerstone objective to maintain functionality of the spent fuel pool system. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance because the finding only affected the barrier function of the spent fuel pool. The inspectors determined that the finding has a human performance crosscutting aspect in the area associated with work control, because Wolf Creek did not coordinate work activities among separate groups, assess the impact of these concurrent evolutions, or track the alignment of the fuel pool clean-up system H.3.b] (Section 1R04).

REPORT DETAILS

Summary of Plant Status The plant started the inspection period at 100 percent rated thermal power. On August 8, 2008, reactor power was reduced to approximately 98 percent and the main generator was reduced to 90 percent for repair of a secondary plant check valve. Wolf Creek was returned to full power on August 10 following repairs. On September 14, reactor power was reduced to approximately 99 percent and the main generator was reduced to 80 percent due to loss of the 345kV Rosehill offsite power line. Wolf Creek was returned to full power the following day after restoration of the power line. Wolf Creek operated at full power for the remainder of the inspection period.

REACTOR SAFETY

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

1R04 Equipment Alignment

.1 Partial Walkdowns

a. Inspection Scope

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

• July 15, 2008, 345kV switchyard alignment

• July 24, 2008, Auxiliary feedwater

• May 2, 2008, Spent fuel pool cooling and cleanup systems The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system, and therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Safety Analysis Report (USAR), Technical Specification requirements, administrative Technical Specifications, outstanding work orders, condition reports (CR), 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. Documents reviewed are listed in the attachment. These activities constitute completion of three partial system walkdown samples as defined by Inspection Procedure

71111.04 - 05.

b. Findings

Introduction.

A Green self-revealing NCV of 10 CFR Part 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified after the licensee utilized two incompatible procedures simultaneously resulting in the inadvertent partial draining of the spent fuel pool. Consequently, approximately 6400 gallons of water was pumped from the spent fuel pool to the refueling water storage tank.

Description.

On May 1, 2008, the licensee was performing Procedure SYS EC-200, "Changing Level in the Spent Fuel Pool or Refueling Pool," Revision 43a, to drain the spent fuel pool to the residual water storage tank. The first step in this procedure is to ensure that the fuel pool cleanup system is shutdown per Procedure SYS EC-320, "Fuel Pool Cooling and Cleanup System Shutdown," Revision 9. With the fuel pool cleanup system secured, the licensee commenced draining spent fuel pool to the residual water storage tank. Following draining, but prior to restoring from the spent fuel pool draining lineup, the licensee re-entered Procedure SYS EC-200 to secure the fuel pool cleanup system in order to drain system containment penetrations. This resulted in a valve alignment that allowed the spent fuel pool cooling pumps to discharge to the residual water storage tank, which caused an unintended lowering of spent fuel pool level from

+2.2 inches to -3.8 inches. After approximately 27 minutes, the operators identified an unexpected increase in residual water storage tank level and took actions to stop the draining from the spent fuel pool. High decay heat fuel assemblies from Refueling Outage 16 were in the spent fuel pool at the time. The water transfer resulted in a 2 degree increase in spent fuel pool temperature.

This issue was entered into the corrective action program which identified the simultaneous performance of the incompatible procedures as the cause of the draining. Wolf Creek's root cause determination identified that the outage control center was not tracking the alignment of the spent fuel pool cleanup system. The inspectors reviewed CRs that identified previous instances of poor coordination of work resulting in a similar lowering of spent fuel pool level. The licensee's root cause for this event was poor implementation of corrective actions from previous occurrences. These corrective actions were to ensure the coordination among work groups and tracking of activities. Both of the incompatible procedures are likely to be performed during the same time period (a refueling outage); however, neither procedure identifies any incompatibilities with the other.

Analysis.

Failure to prevent spent fuel pool draining due to simultaneous performance of incompatible Procedures SYS EC-200 and SYS EC-320 is considered a performance deficiency. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. This finding is more than minor because it impacted the barrier integrity cornerstone attribute of configuration control and affected the cornerstone objective to maintain functionality of the spent fuel pool system. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance because the finding only affected the barrier function of the spent fuel pool. The inspectors determined that this finding has a human performance crosscutting aspect in the area associated with work control, because Wolf Creek did not coordinate work activities among separate groups, assess the impact of these concurrent evolutions, or track the alignment of the fuel pool cleanup system H.3.b].

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion V, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances. Contrary to the above, on May 1, 2008, Procedures SYS EC-200 and SYS EC-320 were not appropriate to the circumstances because neither procedure indicated incompatibility with the other despite the likelihood of simultaneous performance during a refueling outage and resulted in partial draining of the spent fuel pool. Because this finding is of very low safety significance and was entered into the licensee's corrective action program as CR 2008-002035, this violation is being treated as a noncited violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000482-2008004-01, incompatible procedures result in 6400 gallon drain of spent fuel pool.

.2 Complete Walkdown

a. Inspection Scope

The inspectors:

(1) reviewed plant procedures, drawings, the USAR, Technical Specifications, and vendor manuals to determine the correct alignment of the system listed below;

(2) reviewed outstanding design issues, operator work arounds, and corrective action program documents to determine if open issues affected the functionality of the system; and

(3) verified that the licensee was identifying and resolving equipment alignment problems.

• July 14, 2008, Steam turbine-driven auxiliary feedwater system Documents reviewed by the inspectors are listed in the attachment.

These activities constitute completion of one complete system walkdown sample as defined by Inspection Procedure

71111.04 - 05.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

Routine Resident Inspector Tours (71111.05Q)

a. Inspection Scope

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

• July 14, 2008, Circulating water screen house

• July 16, 2008, 2026' elevation of the fuel building

• July 18, 2008, 2018' elevation of the control building

• July 24, 2008, Control building safety-related battery rooms

• September 5, 2008, Engineered safety features transformer yard The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and had implemented adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensee's fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plant's individual plant examination of external events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plant's 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 licensee's corrective action program. Documents reviewed by the inspectors are listed in the attachment. These activities constitute completion of five quarterly fire protection inspection samples as defined by Inspection Procedure

71111.05 - 05.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

Inspection Scope The inspectors reviewed licensee programs, verified performance against industry standards, and reviewed critical operating parameters and maintenance records for the heat exchanger listed below. The inspectors verified that:

(1) performance tests were satisfactorily conducted for heat exchangers/heat sinks and reviewed for problems or errors;

(2) the licensee utilized the periodic maintenance method outlined in EPRI NP-7552, "Heat Exchanger Performance Monitoring Guidelines;"

(3) the licensee properly utilized biofouling controls;

(4) the licensee's heat exchanger inspections adequately assessed the state of cleanliness of their tubes; and

(5) the heat exchanger was correctly categorized under the maintenance rule.

• August 18-19, 2008, Essential service water flushing of auxiliary feedwater room coolers and piping These activities constitute completion of one heat sink inspection sample as defined by IP

R11 Licensed Operator Requalification Program

Resident Inspector Quarterly Review (71111.11Q)

a. Inspection Scope

The inspectors observed testing and training of senior reactor operators and reactor operators to identify deficiencies and discrepancies in the training, to assess operator performance, and to assess the evaluator's critique. The training scenario involved:

• September 5, 2008, Degraded grid voltage followed by station blackout Documents reviewed by the inspectors are listed in the attachment. These activities constitute one quarterly licensed-operator requalification program sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

Routine Quarterly Evaluations (71111.12Q)

a. Inspection Scope

The inspectors reviewed the maintenance activities listed below to:

(1) verify the appropriate handling of structure, system, and component performance or condition problems;

(2) verify the appropriate handling of degraded structure, system, and component functional performance;

(3) evaluate the role of work practices and common cause problems; and

(4) evaluate the handling of structure, system, and component issues reviewed under the requirements of the maintenance rule, 10 CFR Part 50, Appendix B, and Technical Specifications.

• July 18, 2008, Failure of pressurizer heater 480V breaker Documents reviewed by the inspectors are listed in the attachment. These activities constitute completion of one quarterly maintenance effectiveness sample as defined in Inspection Procedure

71111.12 - 05.

b. Findings

No findings of significance were identified.

R13 Maintenance Risk Assessments and Emergent Work Control

1. Risk Assessment and Management of Risk

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

• July 19, 2008, Weekly maintenance planning risk assessment

• August 13, 2008, Yellow risk during planned maintenance on Emergency Diesel Generator B and essential service water Pump B

• September 16, 2008, Unplanned red risk due to simultaneous maintenance on fire-driven diesel pump and essential service water Valve EF-HV52 These activities were selected based on their potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed Technical Specification requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. 2. Emergent Work Control

a. Inspection Scope

For the emergent work activities listed below, the inspectors:

(1) verified that the licensee performed actions to minimize the probability of initiating events and maintained the functional capability of mitigating systems and barrier integrity systems;

(2) verified that emergent work-related activities such as troubleshooting, work planning/scheduling, establishing plant conditions, aligning equipment, tagging, temporary modifications, and equipment restoration did not place the plant in an unacceptable configuration; and

(3) reviewed the corrective action program to determine if the licensee identified and corrected risk assessment and emergent work control problems.

• August 8, 2008, Emergent work on Valve AF-V76

• September 22, 2008, Emergent work on steam generator relief Valve AB-PV4

• May 12, 2008, Emergent work on seat leakage for Valve EP-8818D Documents reviewed by the inspectors are listed in the attachment.

These activities constitute completion of six maintenance risk assessments and emergent work control inspection samples as defined by Inspection Procedure

71111.13 - 05.

b. Findings

Introduction.

Inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," when Wolf Creek mechanically agitated the side of check Valve EP8818D such that the body of the valve was dented numerous times.

Description.

From May 7 to 14, 2008, Wolf Creek was performing evolutions to quantify reactor coolant system leakage past check valves into the safety injection and accumulator tank systems. Leakage from Accumulator D ranged from over 5 percent of tank level per hour to approximately 0.1 percent of tank level per hour. The Technical Specification limit for these pressure isolation valves is 3.0 gpm or less at reactor coolant system normal operating pressure. In particular, Valve EP8818D is down stream of residual heat removal Pump B and the Accumulator D injection line. This 6-inch swing-type check valve prevents reverse flow from Accumulator D to residual heat removal Pump B and opens on forward differential pressure during residual heat removal injection. Check Valve EP8818D failed the acceptance criteria in Procedure STS PE-19E on May 11 and was suspected of allowing leakage from Accumulator D. On May 12, 2008, the Wolf Creek Fix It Now team used a hammer to strike Valve EP8818D and attempt to seat the check valve and reduce the leak. This mechanical agitation was not successful. Wolf Creek's subsequent operability determination stated, in part, that no leakage was observed from the surface of the valve body and that the dents were not deep enough to cause a structural integrity concern. On May 13, 2008, Wolf Creek utilized Procedure TMP 08-018 which started residual heat removal Pump B and flushed water through Valve EP8818D three times in an attempt to remove any debris or boron crystals that may have been preventing the valve from seating. On May 13, 2008, Wolf Creek performed a dye penetrant test on the surface of the valve body for 14 dents on Valve EP8818D and did not find any induced surface cracks. On May 14, 2008, Wolf Creek used an auxiliary pump under Procedures STS PE-19E and STS PE-19F to pressurize the piping between Valve EP8818D and the reactor coolant system check valves, which successfully seated the valve. These procedures were used seven times between May 7 and 14, 2008, to seat Valve EP8818D and other check valves.

The inspectors questioned the mechanical agitation practice. Wolf Creek's engineering evaluation stated, in part, that the mechanical agitation was an acceptable practice, but did not reference any standard. The inspectors could not locate a standard that allowed striking or peening of nuclear grade components as an acceptable maintenance practice. The engineering evaluation also stated, in part, that the striking and denting of the check valve would cold work the surface metal of the valve body and make it stronger. The inspectors found that no engineering evaluation was performed modifying the hardness or strength of the cast austenitic stainless steel prior to the mechanical agitation. The inspectors reviewed design specifications for the valve and found that the body was approximately 1.5 to 3 inches in thickness and that it would be difficult for a person to swing a hammer with enough force to dent the valve internals, regardless of any added strength from cold working. Wolf Creek did not perform any inspection of the welds that attach the valve body to the 6-inch piping. The inspectors reviewed, Troubleshoot, Work Order 08-306327-001, that implemented this practice which used two steel hammers and one brass hammer. This work order did not contain any instructions other than stating that 'rework' may be needed. The inspectors reviewed Wolf Creek Procedure AI 16C-006, "Troubleshooting," Revision 1, and found that it does not contain any such allowance. During an interview, the senior reactor operator who approved the work to commence stated that he did not fully understand what would be done to the valve and that, in retrospect, he would not have allowed the practice.

Analysis.

The inspectors determined that failure to utilize work instructions appropriate to the circumstances and properly evaluate the effects was a performance deficiency. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. The inspectors determined that this issue is more than minor because it could become a more safety-significant concern if the cold working or peening practice is continued. The inspectors determined that the finding was not appropriate for evaluation under Inspection Manual Chapter 0609, Attachment 4. The inspectors applied Inspection Manual Chapter 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria." As a bounding case, if Valve EP8818D is assumed to be damaged and the valve failed to open, residual heat removal Pump B would be unable to inject into reactor coolant system cold leg Number 4 during Mode 3 after the outage. The inspectors did not identify any prohibitions against this practice; however, no additional instances were identified. Additionally, Valve EP8818D opened during flushing and the valve body has substantial wall thickness to resist the strikes. Wolf Creek's dye penetrant examination of the valve body dents did not reveal any cracks. The inspectors did not identify any level of degradation of the valve's ability to open and the valve is considered operable. The inspectors presented these qualitative attributes to NRC management and the finding was determined to be of very low safety significance. The inspectors determined that the finding has a problem identification and resolution crosscutting aspect in the area associated with the corrective action program because the licensee failed to evaluate the problem of seat leakage such that the resolution (a hammer) addressed any possible cause of valve seat leakage P.1(c).

Enforcement.

Part 50 of Title10 of the Code of Federal Regulations, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by documented instructions or procedures of a type appropriate to the circumstances, accomplished in accordance with those instructions or procedures, and contain acceptance criteria to demonstrate that the activity was successfully accomplished. Contrary to the above, on May 12, 2008, Wolf Creek utilized Troubleshoot, Work Order 08-306327-001, which only contained an instruction to "rework" the valve, and workers struck the body of safety-related Valve EP8818D with a hammer several times such that the valve body had been dented. Specifically, no acceptable method of cold working or peening metal surfaces of nuclear grade components was utilized by Wolf Creek, and Procedure AI 16C-006, "Troubleshooting," Revision 1, does not contain any such allowance. Because the finding is of very low safety significance and has been entered into the corrective action program as CR 2008-002284, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000482/2008004-02, use of hammer to reduce accumulator check valve leakage.

R15 Operability Evaluations

a. Inspection Scope

The inspectors:

(1) reviewed plant status documents such as operator shift logs, emergent work documentation, deferred modifications, and standing orders to determine if an operability evaluation was warranted for degraded components;

(2) referred to the USAR and design basis documents to review the technical adequacy of licensee operability evaluations;

(3) evaluated compensatory measures associated with operability evaluations;

(4) determined degraded component impact on any Technical Specifications;

(5) used the significance determination process to evaluate the risk significance of degraded or inoperable equipment; and

(6) verified that the licensee has identified and implemented appropriate corrective actions associated with degraded components.

• July 9, 2008, Bounding stresses for reactor vessel cold leg nozzle welds

• July 24, 2008, Safety-related Battery NK-13 after failed surveillance

• September 5, 2008, Spent fuel pool room Cooler B flange gap exceeds limit

• August 20, 2008, Containment isolation Valve SJ-HV6 dual indication

• September 26, 2008, Essential service water, residual heat removal, and containment spray submerged 4160V cables Documents reviewed by the inspectors are listed in the attachment. These activities constitute completion of five operability evaluation inspection samples as defined in Inspection Procedure

71111.15 - 05.

b. Findings

Introduction.

The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," because Wolf Creek failed to adequately demonstrate that submerged 4160V safety related cables are qualified for such service, and that they will remain operable, although the cables are presently operable.

Description.

During an inspection for license renewal, manhole covers were removed to inspect essential service water cable trays. The cables were found to be submerged. Wolf Creek subsequently initiated CR 2007-3567. The license renewal team addressed the issue with regard to the extended period of operation, after 40 years. The aspect of continued operability for the current 40-year license was referred to the resident inspectors and the Office of Nuclear Reactor Regulation's Electrical Engineering Branch. Wolf Creek's operability evaluation was based on the evaluation from CR 2007-3567 and was entered into the control room log on September 13, 2007. The evaluation relied on NEI White Paper 06-05, "Medium Voltage Underground Cable White Paper," and Work Order 06-285857-000. This work order consisted of an email from the cable vendor, Kerite, stating that the cable insulation is a brown ethylene propylene rubber which had no history of inservice failures. Wolf Creek also cited EPRI TR-103834-P1-2, "Effects of Moisture on the Life of Power Plant Cables," which used moisture absorption following Arrhenius models to extrapolate the remaining life of the cable. Wolf Creek could not produce manufacturer documentation stating that the subject cables were designed for continuous submergence. The inspectors provided this information to the Electrical Engineering Branch for their review.

On September 3, 2008, the Electrical Engineering Branch issued "Wolf Creek Generating Station - Request for Additional Information Related to Submerged Safety-Related Medium Voltage Power Cables (TAC No. MD7339)," (see ADAMS Accession No. ML082410119). Although the cables are presently operable, based on no failures, the NRC is concerned with the continued operability of these cables. The Electrical Engineering Branch questioned 10 aspects of Wolf Creek's position that was documented in CR 2007-3567. These included, in part, the application of the Arrhenius models, operating experience that includes failure of high temperature Kerite cables, the veracity of vendor test data, the difference in water absorption between 100 percent humidity and submergence, the applicability of factory tests to present cable conditions, cable inspection practices, and the application of compensatory measures stated in NEI 06-05. Based on these questions, the Electrical Engineering Branch stated that Wolf Creek has not demonstrated that the subject safety-related cables are qualified for submerged conditions for the current license period.

The inspectors noted that in addition to the cable vaults, that residual heat removal Train A and Containment Spray Train A electrical penetrations in the auxiliary building (1968') are leaking water. These cable conduits are not accessible via manhole.

Wolf Creek previously received a Green finding in Inspection Report 2006010 for failing to establish appropriate procedures for the inspection of buried safety-related electrical cables. Specifically, Wolf Creek did not determine if buried safety-related cables were subject to the degradation described in NRC Information Notice 2002-12, dated March 21, 2008, "Submerged Safety-Related Electrical Cables," or implement any inspections of those submerged cables. Wolf Creek has subsequently written Work Requests 08-311357-000 and 08-311356-000 (and associated work orders) to dewater and inspect the cable vaults.

Analysis.

Failure to perform an operability evaluation that ensured continued operability was considered a performance deficiency. The inspectors determined that this finding was more than minor using Inspection Manual Chapter 0612 Appendix E, Example 3.j, because the NRC was able to show that Wolf Creek's operability evaluation needed significant change to demonstrate continued operability. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance. Specifically, the deficiency did not result in the present loss of operability or functionality and did not represent a risk significant external event such as flooding. The inspectors determined that this finding has a problem identification and resolution crosscutting aspect in the area associated with the corrective action program. Despite several opportunities since 2002, Wolf Creek failed to perform a thorough evaluation for continued operability of submerged safety-related cables to e continued nuclear safety P.1(c).

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to ensure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, from March 21, 2002, to September 3, 2008, Wolf Creek had not adequately evaluated design basis information to identify the acceptable environmental conditions that ensure the safety-related cables are qualified for submerged service. Specifically, Wolf Creek's analytical methods were insufficient to justify continued operability. However, because of the very low safety significance and because the issue was entered into Wolf Creek's corrective action program as CR 2008-005073, this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy:

NCV 05000482/2008004-03, failure to adequately evaluate submerged safety-related cables.

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors selected the below listed postmaintenance test activities of risk significant systems or components. For each item, the inspectors:

(1) reviewed the applicable licensing basis and/or design-basis documents to determine the safety functions;

(2) evaluated the safety functions that may have been affected by the maintenance activity; and

(3) reviewed the test procedure to ensure it adequately tested the safety function that may have been affected. The inspectors either witnessed or reviewed test data to verify that acceptance criteria were met, plant impacts were evaluated, test equipment was calibrated, procedures were followed, jumpers were properly controlled, the test data results were complete and accurate, the test equipment was removed, the system was properly realigned, and deficiencies during testing were documented. The inspectors also reviewed the USAR and corrective action program to determine if the licensee identified and corrected problems related to postmaintenance testing.

• July 18, 2008, Battery NK13 following equalizing charge

• August 18, 2008, Emergency Diesel Generator B postmaintenance test run

• August 13, 2008, Essential Service water Valve EF-HV98 replacement and disc and seat inspection

• August 13, 2008, Essential service water Valve EF-PDV020 replacement and disc seat inspection

• August 13, 2008 Replacement of Emergency Diesel Generator B air start distributor filter

• September 25, 2008, Replacement of Emergency Diesel Generator A and B lube oil heat exchangers Documents reviewed by the inspectors are listed in the attachment.

These activities constitute completion of six postmaintenance testing inspection samples as defined in Inspection Procedure

71111.19 - 05.

b. Findings

Introduction.

On April 7, 2008, inspectors identified a Green, NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," due to an approximately 10-15 gpm leak on the Emergency Diesel Generator B lubricating oil heat exchanger cover plate. The inspectors found that the work order to assemble the heat exchangers was inadequate.

Description.

On April 7, 2008, maintenance in the Wolf Creek switchyard caused a loss of offsite power, resulting in declaration of a Notice of Unusual Event. Emergency Diesel Generator A was already out of service. The reactor was defueled. Emergency Diesel Generator B started as designed but the lubricating oil heat exchanger experienced a leak. When the service water system shut down and was restarted by Emergency Diesel Generator B, a pressure pulse was experienced throughout the essential service water system which caused a portion of the heat exchanger gasket to be ejected. This created a leak between 10 and 15 gpm. A Wolf Creek mechanic stopped the leakage during the Notice of Unusual Event by using a box end wrench to tightened the cover plate bolts. A postmaintenance test prior to the loss of offsite power did not reveal any cover plate leakage. A subsequent evaluation by engineering titled "EKJ04A and EKJ04B Cover Assembly Gasket," found these heat exchangers were newly installed during the current refueling outage and the leak on Emergency Diesel Generator B occurred during the loss of offsite power event. The system engineer requested maintenance to determine the as-found bolt torque values on the Emergency Diesel Generator A lube oil heat exchanger to help determine the cause of the leak. It is not clear that the fasteners were ever checked for proper torque prior to onsite installation. This check was done under Work Order 08-305289-000. The highest torque value reported was 45 ft-lbs and the lowest was 10 ft-lbs. Four bolts were found "loose." The required torque was 195 to 205 ft-lbs. The Engineering evaluation stated that because no documentation was found that torqued the bolts after receipt of the coolers and the bolts on EKJ04A were found not properly torqued, it is reasonable to assume that the bolts on EKJ04B were similarly not torqued as required by site procedures and that this lack of proper torque on the bolts was a contributing cause for the leakage.

The inspectors also reviewed CR 2008-1660 which concluded that the workers incorrectly marked key Step 1.7.23 'N/A' which directed the torquing of the cover plate that leaked. Mechanics marked the step 'N/A' as they believed it not necessary because it included instructions to install a new gasket if the cover was removed in prior steps. Since the cover plate was not removed during prior steps, the mechanics did not torque the cover plate nuts. Essentially, CR 2008-1660 concluded that since the work order did not state "as required," that the mechanics did not have the latitude to mark the step 'N/A' without a change to the work order per the work control process. The inspectors found that Step 1.7.23 was at best confusing and a missed opportunity to torque the cover plate.

Subsequently, the inspectors reviewed the heat exchanger vendor manual which directed that all gasketed connections be torqued after filling and heating of the heat exchanger. These steps were not incorporated into any work order and this aspect was not identified by Wolf Creek in CR 2008-1660. The inspectors found this to be a significant causal factor.

Analysis.

Wolf Creek's failure to ensure that the configuration of the emergency diesel generator lube oil heat exchangers was per plant design was considered a performance deficiency. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. The inspectors determined that this finding was more than minor because it affected the procedure quality attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that to responds to initiating events and prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance. Specifically, the deficiency did not result in the loss of operability or functionality and did not represent a risk significant external event such as flooding. The inspectors determined that the finding has a human performance crosscutting aspect in the area associated with resources. Specifically, Wolf Creek did not ensure that Work Order 08-305289-000 was adequate to ensure nuclear safety by including vendor instructions or acceptance criteria for both emergency diesel generator lube oil heat exchanger cover plates H.2(c).

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, and drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, on April 7, 2008, Wolf Creek failed to ensure that the work order for installation of each emergency diesel generator lube oil heat exchanger cover plate included the tightening of the nuts to prevent a loss of essential service water. However, because of the very low safety significance and because the issue was entered into Wolf Creek's corrective action program as CR 2008-004982, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000482/2008004-04, emergency diesel generator lube oil heat exchanger leak due to water hammer.

1R22 Surveillance Testing

.1 Routine Surveillance Testing

a. Inspection Scope

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

• July 14, 2008, Diesel-driven fire pump run and monthly fuel oil check

• July 15, 2008, NB02 4kV safety bus loss of voltage and degraded volt trip actuating device operational test

• July 16, 2008, Emergency Diesel Generator B hot restart surveillance

• July 17, 2008, Containment spray Train B noncondensable gas void check

• May 7 and 8, 2008, Stroke-time testing of pressurizer power operated relief Valves 455A and 456A

• April 17, 2008, Residual heat removal full flow test and accumulator/safety injection check valve test

• April 13, 2008, STS PE-19, periodic verification of motor-operated valves The inspectors observed inplant activities and reviewed procedures and associated records to determine whether: any preconditioning occurred; effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing; acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis; plant equipment calibration was correct, accurate, and properly documented; as-left setpoints were within required ranges; the calibration frequency was in accordance with Technical Specifications, the USAR, procedures, and applicable commitments; measuring and test equipment calibration was current; test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied; test frequencies met Technical Specification requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used; test data and results were accurate, complete, within limits, and valid; test equipment was removed after testing; where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable; where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure; where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished; prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test; equipment was returned to a position or status required to support the performance of the safety functions; and all problems identified during the testing were appropriately documented and dispositioned in the corrective action program.

Documents reviewed by the inspectors are listed in the attachment. These activities constitute completion of seven routine samples as defined in Inspection Procedure

71111.22 - 05.

b. Findings

No findings of significance were identified.

.2 Inservice Testing Surveillance

a. Inspection Scope

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

• July 24, 2008, Centrifugal charging Pump B inservice test Documents reviewed by the inspectors are listed in the attachment. These activities constitute completion of one inservice testing sample as defined in Inspection Procedure

71111.22 - 05.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

The drill listed below contributed to drill/exercise performance and emergency response organization performance indicators. The inspectors:

(1) observed the training evolution to identify any weaknesses and deficiencies in classification, notification, and protective action requirements development activities;

(2) compared the identified weaknesses and deficiencies against licensee identified findings to determine whether the licensee is properly identifying failures; and

(3) determined whether licensee performance is in accordance with the guidance of the NEI 99-02 document's acceptance criteria.

• September 18, 2008, Evacuation of the control room and main steam line break Documents reviewed by the inspectors are listed in the attachment. These activities constitute completion of one sample as defined in Inspection Procedure

71114.06 - 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 Items Entered Into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensee's 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. Attributes reviewed included: the complete and accurate identification of the problem; that timeliness was commensurate with the safety significance; that evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrences reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue. Minor issues entered into the licensee's corrective action program as a result 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 followup, the inspectors performed a daily screening of items entered into the licensee's corrective action program. This review was accomplished through inspections of the station's daily CR packages. These daily reviews were performed by procedure as part of the inspectors' daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings of significance were identified.

.3 Selected Issue Followup Inspection

a. Inspection Scope

During a review of items entered in the licensee's corrective action program, the inspectors selected the corrective action report listed below for a more in-depth review. The inspectors considered the following during the review of the licensee's actions:

(1) complete and accurate identification of the problem in a timely manner;

(2) evaluation and disposition of operability/reportability issues;

(3) consideration of extent of condition, generic implications, common cause, and previous occurrences;

(4) classification and prioritization of the resolution of the problem;

(5) identification of root and contributing causes of the problem;

(6) identification of corrective actions; and

(7) completion of corrective actions in a timely manner.

• April 7, 2008, Effect of water hammer on control room air conditioning Unit B

• July 26, 2008, Draining of spent fuel pool due to inadequate valve closure Documents reviewed by the inspectors are listed in the attachment. These activities constitute completion of two in-depth problem identification and resolution samples as defined in Inspection Procedure

71152 - 05.

b. Findings

===.1

=

A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," occurred when the April 7, 2008, loss of offsite power caused the service water and essential service water pumps to shut down and restart, resulting in a plant water hammer that caused an approximately 80 gpm essential service water leak from the control room air conditioning Condenser B.

Description.

The essential service water leak from the control room air conditioning condenser was determined to have been caused by gasket failure. The gasket failure was caused by a water hammer which exceeded the design yield strength of the studs which held the endbell and gasket in place. The studs stretched, causing the gasket to fail and causing essential service water leakage. This challenged the operability of the control room air conditioning Train B and the essential service water system by introducing an unaccounted for leak path from essential service water inventory. Control room air conditioning Train A did not start because Emergency Diesel Generator A was not in service at the time of the loss of offsite power. The calculated limit for leakage from the essential service water system was approximately 140 gpm.

Wolf Creek engineering evaluated this event and found that depending on the bolt stress already induced from initial torque and the extra stress induced from the water hammer, it is likely that the yield strength of the studs was exceeded which allowed the studs to stretch and relieve the gasket stress holding the gasket in place. Wolf Creek found that similar leakage at this joint has occurred in the past upon loss of power and subsequent start of the emergency diesel generator and was documented in Performance Improvement Request 2004-2684. At that time, it was concluded that leakage did not prevent fulfillment of the safety function and that the gasket joint connection design might be inadequate, and Performance Improvement Request 2004-2683 was to address the gasket joint design. However, Performance Improvement Request 2004-2683 did not address the connection or change the studs to a stronger material.

Analysis.

Wolf Creek's operation of the control room air conditioning/essential service water system outside the design limits of the heat exchanger studs was determined to be a performance deficiency. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. The finding was determined to be more than minor because it impacted the mitigating systems cornerstone objective of ensuring the availability, reliability, and operability of systems that respond to initiating events. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance because it did not cause the loss of safety function and did not impact risk for external events. The inspectors determined that the finding has a problem identification and resolution crosscutting aspect in the area associated with the corrective action program. Specifically, Wolf Creek identified that the heat exchanger joint might be inadequate but failed to perform any subsequent corrective action P.1.d].

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulation, Appendix B, Criterion III, "Design Control," requires, in part, that measures be established to ensure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structure, system and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, from October 11, 2004, to April 7, 2008, the licensee failed to control component design specifications on control room air conditioning Condenser B such that the yield strength of the condenser studs would not be exceeded during accident conditions which could affect operability of essential service water and control room heating, ventilation and air conditioning. Because this finding is of very low safety significance and was entered into the licensee's corrective action program as CR 2008-001450, this violation is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 5000482/2008004-05, water hammer caused by loss of offsite power exceeds heat exchanger bolt yield strength.

===.2

=

A Green self-revealing NCV of Technical Specification 5.4.1.a was identified for the failure to close Valve EC-V025 during a lineup to recirculate the residual water storage tank through the spent fuel pool cleanup system. As a result, approximately 1500 gallons of water was inadvertently transferred from the spent fuel pool to the residual water storage tank.

Description.

On July 26, 2008, the residual water storage tank was placed into recirculation through the spent fuel pool cleanup system using Procedure SYS EC-121, "Recirculation of the RWST Through the Fuel Pool Cleanup System," Revision 13, for maintenance purposes. Approximately 5 minutes after placing the system into recirculation, the nonlicensed plant equipment operator second guessed his closing of Valve EC-V025. Valves EC-V025 and -V033 had been known to be difficult to operate. After recognizing that ECV-025 was not fully closed, the operator made a second attempt to close the valve and was successful in closing it. It was discovered that the spent fuel pool level had decreased approximately one inch since placing the residual water storage tank into recirculation. These valves are used to isolate the spent fuel pool cooling heat exchanger from the spent fuel pool cleanup system during recirculation of the residual water storage tank. These two systems were cross-connected for approximately 5 minutes, which resulted in approximately 1500 gallons of spent fuel pool water being transferred to the residual water storage tank.

The inspectors documented a similar finding in Integrated Inspection Report 2006004. One of the corrective actions was that the valves would be rebuilt. The licensee noted that these manually-operated valves were historically difficult to operate; however, preventive maintenance on the valves was deferred until October 6, 2008.

Analysis.

Failure to completely close Valve EC-V025 in accordance with the procedure was considered to be a performance deficiency. This finding is more than minor because it is associated with the barrier integrity cornerstone attribute of configuration control and affected the cornerstone objective to maintain functionality of the spent fuel pool system. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance because the finding only affected the barrier function of the spent fuel pool. The inspectors determined that the finding has a problem identification and resolution crosscutting aspect in the area associated with the corrective action program because Wolf Creek did not take appropriate corrective actions to address the adverse trend in manual valve stem friction in a timely manner, commensurate with its safety significance and complexity P.1(d).

Enforcement.

Technical Specification 5.4.1.a, "Procedures," requires that written procedures be established, implemented and maintained covering the activities specified in Appendix A, "Typical Procedures for Pressurized Water Reactors," of Regulatory Guide 1.33, "Quality Assurance Program Requirements," February 1978. Appendix A, Item 3.h, requires procedures for spent fuel pool cooling system operation. Procedure SYS EC 121, "Recirculation of the RWST Through the Fuel Pool Cleanup System," Revision 13, Step 6.1.3, requires that Valves EC-V025 and -V033 be closed when placing the refueling water storage tank in recirculation through the spent fuel pool cleanup system. Contrary to the above procedure, on July 26, 2008, an operator failed to completely close Valve EC-V025 resulting in approximately 1500 gallons of water being inadvertently transferred from the spent fuel pool to the residual water storage tank. Because this finding is of very low safety significance and was entered into the licensee's corrective action program as CR 2008-003663, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000482/2008004-06, failure to completely close the spent fuel pool valve resulted in a loss of spent fuel pool water inventory.

4OA3 Followup of Events and Notices of Enforcement Discretion

.1 Inadvertent Loss of Reactor Coolant System Level During Shutdown Conditions

a. Inspection Scope

The inspectors evaluated the adequacy of the evaluation that Wolf Creek performed for an inadvertent loss of reactor coolant system level during shutdown conditions on March 23, 2008. The inspectors reviewed strip charts, plant computer data, the root cause evaluation, reactor coolant system drawings, level calculations, operating experience entered into the corrective action program, and control room logs. The inspectors determined that the evolution involved a potential finding and Wolf Creek elevated the CR to a root cause evaluation.

Documents reviewed by the inspectors are listed in the attachment.

These activities constitute completion of one sample as defined in Inspection Procedure 71153-05.

b. Findings

Introduction.

The inspectors identified a self revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," because of an inadequate vent path for the reactor vessel head. The inadequate vent path resulted in the formation of a 2600 gallon void in the reactor vessel head while the plant was shutdown and depressurized.

Description.

On March 23, 2008, Wolf Creek depressurized the reactor coolant system and drained the pressurizer to a level of 367 inches above the fuel or approximately 23 percent. Reactor coolant system pressure was established at atmospheric pressure, approximately 6-10 psig below the volume control tank pressure. These actions were performed in accordance with plant operating Procedure SYS BB-215, "RCS Drain Down with Fuel in Reactor." The operators had completed Sections 6.1 and 6.2 of the procedure to vent the reactor vessel head to the pressurizer and purge the pressurizer with nitrogen.

In 1997, Wolf Creek's corrective action program established Performance Improvement Report 970187, indicating that the reactor vessel level indicating system was required to be operable and was not a procedural requirement in the stations reactor coolant system drain down procedures. The inspectors reviewed Procedure SYS BB-215, "RCS Drain Down with Fuel in Reactor," and noted the procedure allowed reactor coolant system depressurization without the reactor vessel level indicating system being available and in service. The inspectors found that operators appropriately responded to the loss of reactor coolant system level and opened the charging pump suction valve to the refueling water storage tank. The inspectors reviewed plant computer data from March 23-24, 2008, and confirmed that a nitrogen bubble had formed in the reactor vessel head region following reactor coolant system depressurization. As the gas built up, it forced primary coolant out of the reactor vessel and into the pressurizer. Based on plant computer data, the drop of approximately 17 inches in pressurizer level equated to a maximum void size of 2655 gallons of primary coolant that had transferred to the reactor vessel.

Three days prior, the reactor vessel level indicating system was removed from service for maintenance, and pressurizer level instruments were the only indicators of reactor coolant system inventory. Prior venting of the reactor vessel head had been unsuccessful to prevent gas buildup. The licensee concluded that the permanent hard pipe (attached to the vessel head) portion of the vent rig most likely had existing blockage or a loop seal. The licensee found indirect evidence of a loop seal due to water that came out of the hard pipe at the end of the outage during vacuum filling of the reactor coolant system. However, the licensee could not exclude blockage in the piping. The licensee planned to conduct a more thorough inspection of the hard pipe during the next refueling outage. The inspectors determined that the reactor operator's ability to identify an unsuccessful vent lineup was challenged since the primary means of identifying gas accumulation, the reactor vessel level indication system, was not required to be in service. Wolf Creek entered this issue into the corrective action program to evaluate changes to plant operating procedures and to verify permanent hard pipe portions of the reactor vessel vent rig are not blocked or susceptible to creating loop seals. Excluding the void, time to boil in the reactor coolant system was calculated to be 3.76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> during outage planning.

Analysis.

The inspectors determined that failure to provide an adequate vessel head vent path to prevent gas accumulation in the reactor vessel during depressurized plant operations was a performance deficiency. This issue was reasonably within Wolf Creek's ability to foresee and prevent. The inspectors determined that this finding, which was associated with the initiating events cornerstone, was more than minor because if it was left uncorrected, it would have become a more significant safety concern. Specifically, operating procedures did not require an effective means of monitoring the reactor coolant system level to minimize gas buildup when plant conditions were conducive to this phenomena. The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs." The inspectors determined that Checklist 3 was applicable because the unit was in cold shutdown with the refueling cavity level less than 23 feet. Based upon Appendix G, Attachment 1, Checklist 3, Phase 2, analysis was not needed to characterize the risk significance of this finding because the level of loss was less than 2 feet, did not occur during reduced inventory, and appropriate action was taken regarding the level deviation. The finding was determined to be of very low safety significance based upon the demonstrated availability of mitigation systems and the reactor coolant system cavity inventory.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," requires, in part, that the design basis is correctly translated into specifications, drawings and procedures. Contrary to the above, from December 2, 2003, through March 24, 2008, Wolf Creek had designed and installed a reactor vessel head permanent vent piping modification which failed to adequately vent noncondensable gases resulting in the formation of a 2600 gallon void in the reactor vessel head while the plant was shutdown and depressurized. Because this violation is of very low safety significance and has been entered into Wolf Creek's corrective action program as CR 2008-001032, this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000482/2008004-07, inadequate reactor vessel head vent.

.2 Secondary Plant Power Reduction due to Loss of Rosehill 345kV Line

a. Inspection Scope

On September 14, 2008, Wolf Creek reduced generator load in response to loss of the Rosehill 345kV offsite power line. Control room operators maintained reactor power at approximately 99 percent by taking the steam dump controls to the pressure mode of operation and sending steam to the condenser. The inspectors observed the switchyard line-up, control room command and control, procedure use, unit limitations, and walked down the major components of the secondary side. These activities constitute completion of one sample as defined in Inspection Procedure 71153-05.

b. Findings

No findings of significance were identified.

OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security.

These observations took place during both normal and off-normal plant working hours. These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status reviews and inspection activities.

b. Findings

No findings of significance were identified.

.2 Implementation of Temporary Instruction (TI) 2515/176 - Emergency Diesel Generator Technical Specification Surveillance Requirements Regarding Endurance and Margin Testing

a. Inspection Scope

The objective of TI 2515/176 was to gather information to assess the adequacy of nuclear power plant emergency diesel generator endurance and margin testing as prescribed in plant-specific technical specifications (TS). The inspectors reviewed the licensee's TS, procedures, and calculations and interviewed licensee personnel to complete the TI. The information gathered while completing this TI was forwarded to the Office of Nuclear Reactor Regulation for further review and evaluation.

b. Findings

Introduction.

The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to account for the effect of variation in the emergency diesel generator frequency and voltage calculations.

Description.

The inspectors reviewed Drawing E-11005, "List of Loads Supplied by Emergency Diesel Generator," which supplied the load information used as input for the design analysis results contained in Report M-018-01502, "Engineering Report Wolf Creek NPP 6201kW Diesel Generator Set." Report M-018-01502 was a voltage and frequency study of the emergency diesel generator during analyzed accident conditions. The inspectors noted that the frequency and voltage variation determined in the calculations was based on nominal 60 hertz, 4160V operation of pumps and fans, and did not account for the lower two percent variation allowed by Technical Specification 3.8.1. The inspectors reviewed the USAR and found that Section 8.1.4.3.b states, that at no time during load sequencing shall voltage decrease below 75 percent of 4160V and frequency shall not decrease below 95 percent of 60 hertz. This section is based on Regulatory Guide 1.9. These limits equate to approximately 3120V and 57 hertz respectively. Calculation M-018-01502, states that frequency will decrease to 7.57 hertz and 3199v, respectively, when the essential service water pump is started. The inspectors assumed a steady state starting point of the Technical Specification allowed 58.8 hertz, and/or 4076V and Wolf Creek could not demonstrate that bus frequency and voltage would remain above the USAR limits.

In response to this question, the inspectors held discussions with licensee engineers and operations personnel and the licensee performed an operability evaluation. Wolf Creek agreed that Calculation M-018-01502 did not analyze or demonstrate operability at the lower bound of the Technical Specifications. Using recorded data, Wolf Creek was able to show that the diesel would stay slightly above the 57 hertz limit. Wolf Creek had data from the integrated safeguards testing performed during the most recent refueling outage in March and April 2008. During those tests, all loads are sequenced onto the safety bus with the diesel supplying the bus. During the Emergency Diesel Generator A test, the frequency decreased to 57.16 hertz when the essential service water pump was started. During the Emergency Diesel Generator B test, bus frequency decreased to 57.45 hertz when the essential service water pump was started. Additionally, during an actual loss of offsite power, Emergency Diesel Generator B started and bus frequency decreased to 58.7 hertz. Lastly, Wolf Creek procedures for the engine control Unit 2301A and governor setup have acceptance criteria of 516-518 rpm which equates to 59.9-60.1 hertz. Assuming no setpoint drift, the inspectors concluded that the current setup of the governors can ensure operability. Wolf Creek initiated CR 2008-4312 to recalculate the voltage and frequency response with the lower Technical Specification limit.

Analysis.

The inspectors determined that failure to properly account for the effect of frequency variation on the diesel generator was a performance deficiency. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. The finding was determined to be more than minor because the finding was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, based on current calculations, the failure to account for frequency variations had more than a minimal effect on the outcome of the analysis in that the 57 hertz low frequency limit for the emergency diesel generators would have been exceeded in loss of coolant accident and loss of offsite power scenarios with the assumed loads. This assumes a steady state starting point of the Technical Specification allowed frequency of 58.8 hertz. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined be of very low safety significance because the finding represented a design or qualification deficiency confirmed not to result in loss of operability.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to ensure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, prior to August 29, 2008, the licensee had not adequately translated design basis information into the diesel generator frequency and voltage analysis. Specifically, the analysis providing the information in M-018-01502, "Engineering Report Wolf Creek NPP 6201kW Diesel Generator Set," did not properly account for the Technical Specification allowable diesel generator two percent frequency and voltage reduction. The licensee failed to consider how the frequency and voltage variation could affect the design and licensing basis of the diesels. Because the violation was of very low safety significance and the licensee entered the finding into their corrective action program as CR 2008-04312, this violation is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000482/2008004-08, diesel generator low frequency and voltage variation not considered in calculations.

4OA6 Meetings, Including Exit

On October 3 and 15, 2008, the resident inspectors presented the inspection results to Mr. M. Sunseri, Plant Manager, and other members of the licensee staff. The licensee acknowledged the findings presented.

A-1 Attachment

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

R. A. Muench, President and Chief Executive Officer

M. Sunseri, Vice President Operations and Plant Manager

S. E. Hedges, Vice President Oversight

K. Scherich, Director Engineering

T. East, Manager, Emergency Planning

P. Bedgood, Superintendent, Chemistry/Radiation Protection

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000482/2008004-01 NCV Incompatible procedures result in 6400 gallon drain of spent fuel pool (Section 1RO4)

05000482/2008004-02 NCV Use of hammer to reduce accumulator check valve leakage (Section 1R13)

05000482/2008004-03 NCV Failure to adequately evaluate submerged safety-related cables (Section 1R15)

05000482/2008004-04

NCV Emergency diesel generator lube oil heat exchanger leak due to water hammer (Section 1R19)

05000482/2008004-05

NCV Water hammer caused by loss of offsite power exceeds heat exchanger bolt yield strength (Section 4OA2)

05000482/2008004-06

NCV Failure to completely close the spent fuel pool valve resulted in a loss of spent fuel pool water inventory (Section 4OA2)

05000482/2008004-07

NCV Inadequate reactor vessel head vent (Section 4OA3)

05000482/2008004-08

NCV Diesel generator low frequency and voltage variation not considered in calculations (Section 4OA5)

LIST OF DOCUMENTS REVIEWED

In addition to the documents referred to in the inspection report, the following documents were selected and reviewed by the inspectors to accomplish the objectives and scope of the inspection and to support any findings:

Attachment

Section 1R04: Equipment Alignment

Procedures

ALR 00-128B, TD AFW Start, Revision 8

CKL

AL-120, Auxiliary Feedwater Normal Lineup, Revision 34

SYS

AL-120, Feeding Steam Generators with a Motor Driven or Turbine Driven AFW Pump,

Revision 33

SYS

KJ-121,Diesel Generator Lineup for Auto Ops, Revision 39

Drawings

M-12AB02, Main Steam System, Revision 11 M-12FC02, Auxiliary Feedwater Pump Turbine, Revision 21

Work Orders

08-306201-000 08-305006-001 08-302049-000 08-302360-000

Condition Reports

2008-001996

2008-002941

Miscellaneous

Engineering Disposition, Relocate I/P From The ARVs, ABPV001 Thru 004, Revision 6

KD-7496, One Line Diagram, Revision 33 Wolf Creek Generating Station USAR, Revision 19

Section 1R05: Fire Protection

Procedures

ALR

KC-888, Fire Protection Panel

KC-008 Alarm Response, Revision 15

AP 10-106, Fire Preplans, Revisions 5 and 6 STN

FP-815A, Heat Trip Actuation Device Operational Test, Revision 3

Condition Reports

2007-002929

Work Requests

07-063647

Work Orders

06-284430-000 06-284436-000

Attachment Drawings E-OFO221, Fire Detection/Protection System-Yard Transformer Area EL. 2000'-0", Revision 5

Miscellaneous

Fire Hazard Analysis Fire Area

ESF-YARD, Revision 0 Fire Hazard Analysis Fire Area H-1, Revision 0 Fire Hazard Analysis, Fire Area CST & RWST, Revision 0 M-650-00119, Fenwal Detect-a-Fire Detection and Release Devices Post Fire Safe Shutdown Area Analysis, E-1F9910, Revision 2 Prefire Plan, Auxiliary Building Prefire Plans, Revision 6

Prefire Plan, Fire Protection Water Supply and Hydrant Locations, Revision 0 Wolf Creek Generating Station Individual Plant Examination Summary Report, September 1992

Section 1R11: Operator Requalification

Procedures

AI 21-100, Operations Guidance and Expectations, Revision 8

AP 21-001, Conduct of OPS, Revision 36A

APF 06-02-001, Emergency Action Levels, Revision 8

EDI 23M-050, Monitoring Performance to Criteria and Goals, Revision 3 EMG

CS-01, Loss of all AC Power Recovery without SI Required, Revision 17 EMG

ES-11, Post LOCA Cooldown and Depressurization, Revision 17

Miscellaneous

Operations Requalification Cycle 08-05, Revision 0

Section 1R12: Maintenance Effectiveness

Performance Improvement Requests

2008-003160

Work Requests

07-061766 07-061883 07-061884 07-060117 07-060141

07-060514 07-059846

Work Orders

08-065850

08-303805-000

Condition Reports

2008-003160

Maintenance Rule Maintenance Rule Final Scoping Evaluation

BB-14 - Reactor Coolant System Attachment Maintenance Rule Scoping Evaluation for System

PG-01 - Low Voltage 480V

Miscellaneous

Calculation

AN-99-031, Development of PSA based Reliability Performance Criteria for Maintenance Rule, Revision 0

EDI 23M-050 Attachment B, Functional Failure Determination Checklist

EDI 23M-050, Engineering Desktop Instruction Monitoring Performance to Criteria Goals, Revision 3

INC

C-1000, Calibration of Miscellaneous Components, Revision 7

STS

MT-001A, PG21 Pressurizer Heater Verification, Revision 2

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

Procedures

AIF 22C-006-01, Checklist for Emergent Work, Revision 4,

ALR 00-045C, Acc Tk C Lev HILO, Revision 12A

AP 05-005, Design, Implementation & Configuration Control of Modifications, Revision 12

AP 16E-002, Post Maintenance Testing Development, Revision 6A

AP 21I-001, Temporary Modifications, Revision 8

AP 22C-003, Operational Risk Assessment Program, Revision 11

GEN 00-004, Power Operation, Revision 59 STS

PE-019B, Residual Heat Removal Suction Valve Leak, Revision 16 STS

PE-019E, Reactor Coolant System Isolation Check Valve Leak Test, Revision 23 STS

PE-019F, Individual Leak Test of EPV0010/20/30/40 and EP8818A/B/C/D, Revision 1 SYS

AC-322, MSR 2

nd Stage Reheat Operations, Revision 11

TMP 08-020, ECCS Accumulators Sample Valve Leak Check, Revision 0

Condition Reports

2008-003721

2008-002284

2008-002230

2008-002237

Work Orders

07-300584-000

07-300584-001 07-300584-002 07-292506-000 06-282700-000 07-297055-000 08-306327-016 08-307295-001 99-214533-000 05-272949-000 96-115242-006 96-115242-005 95-100101

Miscellaneous

APF 22C-003-001, Operational Risk Assessment, Revision 0, 7/27-8/2/2008

Drawings

FB-158950, 8" No. 7109U (L-973A) P.S. Tilting Disc Check Valve, Revision 1

Attachment M-12EJ01, P&ID Residual Heat Removal System, Revision 42 M-724-00352, Swing Check Valve

MOD 06001CS990000D0, Revision W05 M-724-00354, Swing Check Valve

MOD 10001CS990000D0, Revision W04

Section 1R15: Operability Evaluations

Calculations

GG-S-001, Evaluation of Misalignment of flange joints - pump room cooler (SGG04A), Revision 0

Drawings

M-12SJ04, Piping and Instrumentation Diagram Nuclear Sampling System, Revision 7 M-13EF12(Q), Piping Isometric Essential Service Water Fuel Building, Revision 5 USAR figure 6.2.4-1, Pages 45a and 67, Containment Penetrations, Revision 13

Procedures

AP 16G-002, Installation of Piping, Revision 2

KMS-4, Table 1, Torque Tables for Raised Face Flanges, Revision 1

MGM

MOOP-08, Torquing Guidelines for Bolted Connections, Revision 11

MPE-BA-013, Charging Individual Battery Cells, Revision 5

MPE-E009Q-03, Inspection and testing of Siemens vacuum circuit breakers

STS

KJ-001A , Integrated D/G and Safeguards Actuation test, Train A performed on November 12, 2003

STS

MT-019, 125VDC Class 1E Quarterly Battery Inspection, Revision 15

Condition Reports

2007-004329

2007-003704

2007-003462

2008-003331 2008-003589

2008-004121

2008-004062

2008-004415 2008-004408

2008-004375

2007-003567

2006-0390 2002-2125

Work Orders

08-308247-001 08-308247-006 08-308247-012 08-309246-002 08-309246-001 08-309568-006 06-286530-000 08-309568-001 08-309568-000 06-286530-000 06-285857-000 07-299577-000 06-284064-000 08-311357-009 08-311356-006 06-284064-000

07-294641-000

Miscellaneous

Change Package

012681, Spent Fuel Pool Room Cooler SGG04A, Revision 0

Attachment USAR Section 9.3.2 WR 08-066708

Section 1R19: Postmaintenance Testing

Procedures

CKL

ZL-005B, B EDG Operating Log, Revision 1, 8-14-08 I-ENG-010 Use of Diesel Engine Analyzer Equipment, Revision 1

STS

MT-019, 125VDC Class 1E quarterly battery inspection, Revision 15

SYS

KJ-124, Post Maintenance Run of Emergency Diesel Generator B,

Work Orders

08-300437900-1 08-308247-006 08-305917001-1 08-306347000

307526000-1

2998510052

299849004

Condition Reports

2007-003712

2007-003825

2007-003912

2007-004117 2008-003396

2008-003732

2007-004117

2008-003912

2008-003925

2008-001450

2008-001481

2008-001660 2008-001782

2008-001851

2008-001898

Miscellaneous

Change Package 12708, Evaluation of ESW Water Hammer Event Due to Loss of Offsite Power, Revision 0

Change Package 12746, Torque of Piping Flanges Between EDG Heat Exchangers, Revision 0

Design Specification for Replacement Emergency Diesel Generator Lube Oil Heat Exchangers for the Wolf Creek Generating Station, Revision 01

Engine Combustion Report - EDG B

ESW MOV disc and seat inspections and Limitorque replacements for valve ESW Train B EFHV0098 Air release valve and EFHV0020 Strainer trash valve

I-ENG-010 Use of Diesel Engine Analyzer Equipment, Revision 1

KKJ01B 2008 Parameter Trends

M-018C-00002, EDG Heat Exchangers Storage, Installation, Operation, and Maintenance Instruction Manual, Revision W02

Purchase Order No.

737566

SYS

KJ-124, Postmaintenance run of EDG B

Attachment Procedures

AP 20E-001, Industry Operating Experience Program, Revision 9 SYS

KJ-123, Post Maintenance Run of Emergency Diesel Generator A, Revision 38 SYS

KJ-200, Inoperable Emergency Diesel, Revision 13

Section 1R22: Surveillance Testing

Procedures

AI 23D-003 MOV Trending and Periodic Verification Program, Revision 1

M-12EN01, Piping and instrumentation diagram containment spray system, Revision 11

STN

EN-003, Containment spray system void monitoring and venting, Revision 1

STN

FP-211, Diesel Fire Pump 1FP01PB Monthly Operation and Fuel Level Check, Revision 16

STS

BB-204, Rector Coolant System Inservice Valve Test, Revision 11

STS

BG-100B, Centrifugal Charging System "B" Train In-service Pump Test, Revision 34

STS

CV-211, Residual Heat Removal Full Flow Test and Accumulator Safety Injection Check Valve Test, Revision 16

STS

IC-013B, Hot restart of EDG NE02, Revision 7

STS

IC-208B, 4kV loss of voltage and degraded voltage TADOT NB02 bus, separation Group 4, Revision 2A

STS

PE-007, Periodic Verification of Motor Operated Valves (EJHV8811A), Revision 3

STS

VT-001 Verification of

OMN-1, MOV Exercise Requirements, Revision 3A

Miscellaneous

ASME Code Case

OMN-1 Alternative Rules for Pre-service and In-service Testing of Certain Electric Motor Operated Valves, OM Code-1995 Subsection ISTC

Technical Position

TP-11, Motor Operated Valve

OMN-1 Code Case Position

USAR Section 15.2, Revision 11

WCOP-02, In-service Testing Program Third Ten Year Interval, Revision 14

Wolf Creek Response to Generic Letter 90-06

Drawings

M-12BB01 P&ID Reactor Coolant System, Revision 25 M-12BB02 P&ID Reactor Coolant System, Revision 16

Attachment M-12BB03 P&ID Reactor Coolant System, Revision 11 M-12BG03 P&ID Reactor Coolant System, Revision 39 M-12EJ01 P&ID Residual Heat Removal System, Revision 42

Condition Reports

2008-001806

2008-003531

2008-003529

2008-003584 2008-003584

Work Orders

07-296486-000 05-279238-000 05-279238-001 05-279238-002 05-279238-003 05-79238-004

36022

Section 1EP6: Drill Evaluation

Procedures

APF 06-002-01, Emergency Action Levels, Revision 22 OFN

RP-17, Control Room not Habitable, Revision 11

Condition Reports

2008-004637

2008-004617

2008-004618

2008-004619 2008-004621

2008-004622

2008-004623

2008-004624 2008-004861

2008-004880

2008-004881

Miscellaneous

September 18, 2008 Drill Evaluation Drill Scenario 08-SA-02

Section 4OA2: Problem Identification and Resolution

Condition Reports

2008-003589

2008-004121

2008-004062

2008-004415 2008-004408

2008-004375

2007-003567

2006-0390

Work Orders

06-284064-000 08-311357-009 08-311356-006 06-284064-000 07-294641-000

Procedures

AP 16G-002, Installation of Piping, Revision 2

KMS-4, Table 1, Torque Tables for Raised Face Flanges, Revision 1 MGM

MOOP-08, Torquing Guidelines for Bolted Connections, Revision 11

Attachment

Section 4OA3: Event Followup

Condition Reports

1997-97087

2008-003053

2008-001032

2008-001059

Procedures

AP 22B-001, Outage Risk Management, Revision 9

FHP 00-001, Operations Outage Checklist, Revision 5

GEN 00-004, Power Operation, Revision 59

GEN 00-009, Refueling, Revision 20

SYS

BB-215, Reactor Coolant System Drain Down with Fuel in the Reactor, revision 20A

Drawings

M-13BB17, Small Pipe Isometric Reactor Coolant Sys. Vessel Level Instr. & Head Vent -

Reactor Bldg, Revision 4

WCRE-03,

TBB 03 Pressurizer, Revision 22

Miscellaneous

Change Package 09953, Reactor Vessel Simplified Head Assembly (SHA), Revisions 9 and 13

Section 4OA5: Other Activities

Calculations

KJ-M-010, Heat Exchanger Analysis, August 28, 2002

KJ-M-011, EKJ03A/B Thermal Performance Calculation, Revision 0

XX-E-006, Demand Load Study, Revision 5

Drawings

E-11005, List of Loads Supplied by Emergency Diesel Generator, Revision 32 E-11005, List of Loads Supplied by Emergency Diesel Generator, Revision 32 modified

Procedures

BD-EMG

ES-11, Post LOCA Cooldown and Depressurization, Revision 17

BD-EMG

ES-11, Post LOCA Cooldown and Depressurization, Revision 14

BD-EMG

ES-11, Post LOCA Cooldown and Depressurization, Revision 7

EMG E-1, Loss of Reactor or Secondary Coolant, Revision 18A I-ENG-024, Data Acquisition, Revision 4 MPE

NE-002, Governor Adjustments for Emergency Diesel Generator NE02, Revision 9 OFN

NB-042, Loss of Offsite Power to NB01 (NB02) With EDG Paralleled, Revision 0 STS

KJ-011A, DG NE01 24hour Run, Revision 19

STS

KJ-011B, DG NE02 24hour Run, Revision 19

Attachment Miscellaneous Colt-Pielstick -

PC2.5V Engine Instructions Engineering Disposition, Risk Impact of EDG Parallel to the Grid for Load Testing, Revision 1

Instructions for Type WNR Voltage Regulator Static Excitation System for Auxiliary Diesel Generator Systems, January 1983

License Amendment No. 101

License Amendment No. 164

M-018B-00001, Instruction Manual for [EDG] Governor, Revision W03

MA 92-010, Hardware Failure Analysis, EDG B Lube Oil Temperature Control Valve Special Report 85-010

Special Report 86-011

Special Report 87-015

Special Report 87-016

Special Report 88-002

Special Report 88-004

Special Report 89-003

Special Report 95-002

Specification No. 10466-M-018(Q), [EDG] Design Specification for the Standardized Nuclear

Unit Power Plant System (SNUPPS), Revision 6

USAR Section 3.1, Revision 1

USAR Section 8.3.1.1.3

USAR Table 6.2.2.3-3, Revision 15

Condition Reports

2007-002683

2007-2734

PIR-98-3464

PIR-2000-0423

PIR-2004-2158

PIR-2005-1044 2007-4117

2007-4719 2007-0368

2008-2718