IR 05000482/2007002

IR 05000482-07-002; 01/01/07 - 04/07/07; Wolf Creek Generating Station; Fire Protection, Followup of Events and Notices of Enforcement Discretion, Access Control to Radiologically Significant Areas

ML071310291
Person / Time
Site:	Wolf Creek
Issue date:	05/11/2007
From:	Vincent Gaddy NRC/RGN-IV/DRP/RPB-B
To:	Muench R Wolf Creek
References
IR-07-002
Download: ML071310291 (46)

Text

May 11, 2007

SUBJECT:

WOLF CREEK GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000482/2007002

Dear Mr. Muench:

On April 7, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Wolf Creek Generating Station. The enclosed integrated report documents the inspection findings which were discussed on April 11, 2007, with Mr. S. E. Hedges and other members of your staff.

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

Inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents six NRC-identified and self-revealing findings of very low safety significance (Green). Five of these findings were determined to involve violations of NRC requirements. The NRC is treating these violations as noncited violations consistent with Section VI.A.1 of the NRC Enforcement Policy because of the very low safety significance and because the findings were entered into your corrective action program. 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.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its enclosure, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRCs document system (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Wolf Creek Nuclear Operating Corporation-2-Should you have any questions concerning this inspection, we will be pleased to discuss them with you.

Sincerely,

/RA/

Vincent G. Gaddy, Chief Project Branch B Division of Reactor Projects Docket: 50-482 License: NPF-42

Enclosure:

NRC Inspection Report 05000482/2007002 w/Attachment: Supplemental Information

REGION IV==

Docket:

50-482 License:

NPF-42 Report:

5000482/2007002 Licensee:

Wolf Creek Nuclear Operating Corporation Facility:

Wolf Creek Generating Station Location:

1550 Oxen Lane NE Burlington, Kansas Dates:

January 1 to April 7, 2007 Inspectors:

S. D. Cochrum, Senior Resident Inspector J. Reynoso, Reactor Inspector C. Long, Resident Inspector B. K. Tharakan, Health Physicist Approved By:

V. G. Gaddy, Chief, Project Branch B

Enclosure-2-

SUMMARY OF FINDINGS

IR 05000482/2007002; 01/01/07 - 04/07/07; Wolf Creek Generating Station; Fire Protection,

Followup of Events and Notices of Enforcement Discretion, Access Control To Radiologically Significant Areas This report covered a 3-month period of inspection by resident and regional inspectors. The inspection identified six Green findings, five of which were 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 review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

Cornerstone: Mitigating Systems

Green.

A self-revealing noncited violation of Technical Specification 5.4.1.d was identified for failure to implement fire protection impairment control permit requirements and compensatory measures when operators incorrectly disabled three fire detectors in the auxiliary building. The detectors in the auxiliary building were disabled without a proper fire impairment control permit and the required compensatory roving hourly fire watch for a period of approximately hours as required by Administrative Procedure AP 10-103, Fire Protection Impairment Control, Revision 21. This issue is captured in the licensees corrective action program.

The failure to implement fire protection impairment control permit requirements and establish compensatory measures for the auxiliary building 1974' level was considered a performance deficiency. The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it involved compensatory measures for the fixed fire protection system and was assigned a low degradation rating since less than 10 percent of the fire detectors in the area were disabled. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to apply appropriate human error prevention techniques such as self and peer-checking prior to removing the fire detectors from service. (Section 1R05)

Cornerstone: Miscellaneous

Green.

A noncited violation of Technical Specification 5.4.1.a was identified for failure to maintain sufficient records (logs) to furnish evidence of events significant to plant safety. On January 26, 2007, electrical maintenance commenced a scheduled replacement of main control board Annunciator Power Supply E1PS5. During the power supply replacement, a loss of 8.7 percent of the annunciators was expected. However, during de-termination of the power supply leads, an unexpected loss of a significant number of the main control board annunciators occurred. Subsequently, due to the large number of annunciator inputs that were lost, the plant computer became overloaded and stopped updating. Based on these indications, the control room operators would need to evaluate emergency action level and Technical Specification requirements. The inspectors discovered during interviews with the operations crew that was on watch during the event, that no information was recorded or kept during the event. Administrative Procedure AP 21-001, Conduct of Operations, Revision 36A, requires operators to make plant log entries of potentially reportable occurrences, entries that could be useful in reconstructing events, and events significant to plant safety. However, the logs were not updated until several hours later based on verbal accounts provided to the oncoming crew. The inspectors noted that the after the fact log entries still provided insufficient data to reconstruct the activities related to the loss of annunciators. This issue is captured in the licensees corrective action program.

The failure to adequately document times and information for the loss of annunciators was considered to be a performance deficiency. This finding was more than minor because it could impact the operators ability to accurately implement emergency action levels and Technical Specification action statements and if left uncorrected, this type of insufficient documentation could become a more significant safety concern. The finding required NRC management review and was determined to be of very low safety significance because the loss of annunciators challenged the emergency action level time requirements but was restored prior to exceeding any emergency action level or Technical Specification action time requirement. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively communicate expectations regarding plant operating log entries in accordance with procedural requirements. (Section 4OA3.1(1))

Green.

A self-revealing finding was identified regarding inadequate engineering drawings used as guidance to replace main control board annunciator power supplies resulting in a loss of all main control board annunciators. During de-termination of the power supply leads, an unexpected loss of a significant number of the annunciators occurred. During the planning review of Work Order 06-280217-003, Replace Power Supply RK045E1PS5, the electricians brought forth a concern about the daisy chaining of the leads associated with the main control board power supplies and not knowing what effect removing a power supply would have on additional annunciators. System engineering reviewed vendor drawings and determined that only the expected annunciators would be lost. The vendor drawings only consisted of discrete wire connections from the power supply to the logic bus and did not show interconnections with any other power supplies. Although, it was acknowledged by system engineering that there were numerous daisy chained connections not shown on the vendor drawings, no further reviews or research was conducted. The licensees root cause analysis determined that the vendor drawings did not show the interconnecting wiring identifying point-to-point connections associated with the main control board power supplies. This issue is captured in the licensees corrective action program.

The failure to maintain drawings technically accurate and reflect the as-built condition of the plant was considered to be a performance deficiency. The finding was more than minor because it impacted the maintenance technicians ability to accurately plan and implement work, resulting in the annunciator system being adversely affected and could be reasonably viewed as a precursor to a significant event. The finding required NRC management review and was determined to be of very low safety significance because the finding did not result in a loss of a system safety function or a loss of risk significant equipment for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This finding has a crosscutting aspect in the human performance area associated with the resources component because the licensee failed to maintain complete, accurate and up-to-date design documentation. (Section 4OA3.1(2))

Cornerstone: Barrier Integrity

Green.

A noncited violation of Technical Specification 5.4.1.a occurred when operators did not take timely action to lower power below the licensed thermal limit of 3565 MWt. During an incore to excore neutron detector calibration, the power level exceeded the limit and the operating crew did not insert negative reactivity until after the neutron detector calibration was complete. During this evolution, the reactor exceeded licensed thermal power of 3565 MWt for approximately 58 minutes, peaking at 3566.5 MWt according to the plant computers 10-minute calorimetric. After the neutron detector calibration was completed, operators added boron to the reactor coolant system to reduce power below 100 percent. Procedure GEN 00-004, Power Operation,

Attachment B, Step B.1.1 states, in part, that exceeding 3565 MWt is permitted only as a result of transients or computer point fluctuations. The inspectors judged that allowing reactor power to ascend above 100 percent for nearly an hour was not a transient. However, operators did not initiate action in accordance with Step B.1.1 when the 10 minute average exceeded 3565 MWt until approximately 40 minutes elapsed. This issue is entered into the corrective action program.

The failure to maintain steady state reactor power at or below the licensed thermal power limit is a performance deficiency. The finding was more than minor because it is associated with the configuration control attribute for the Barrier Integrity Cornerstone; and, it affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radio nuclide releases caused by accidents or events. The finding was of very low safety significance because the fuel cladding barrier was affected but did not affect the reactor coolant system or containment barriers. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee did not ensure that licensed operators used conservative assumptions in their decision making when reactor power increased above the licensed limit for an extended period. (Section 4OA3.2)

Green.

A self-revealing noncited violation of Technical Specification 5.4.1 occurred on February 20, 2007, when a chemistry technician inadvertently removed both containment purge radiation Monitors GTRE22 and GTRE33 from service at the same time. During planned maintenance on the safety-related GTRE33 containment purge radiation monitor, a chemistry technician inadvertently removed the incorrect containment purge radiation monitor from service. After contacting the control room, the shift chemist went to GTRE22 and incorrectly removed the radiation monitor from service. Instrumentation and controls personnel working at GTRE33 informed the shift chemist that the incorrect radiation monitor was removed from service. The shift chemist subsequently returned GTRE22 to service. Technical Specification 3.3.6,

Condition A, was entered for having more than one train inoperable. The containment purge and supply dampers were immediately verified to be closed and remained closed with no containment purge in progress. This issue was entered into the licensees corrective action program.

The inspectors determined that the failure to remove the correct containment radiation monitor from service was a performance deficiency. The finding was more than minor because it is associated with the configuration control attribute for the Barrier Integrity Cornerstone; and it affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radio nuclide releases caused by accidents for events. The finding was of very low safety significance because both trains of the radiation monitor protective functions (i.e., to stop a containment purge on a high radiation signal) were affected but did not result in an actual open pathway in the containment barrier. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the shift chemist failed to apply appropriate human error prevention techniques such as self and peer-checks. (Section 4OA3.3)

Cornerstone: Occupational Radiation Safety

Green.

The inspector reviewed a self-revealing noncited violation of 10 CFR 20.1501(a) because the licensee failed to perform an adequate survey in a high radiation area. On March 7, 2007, a health physics technician performed a survey of Floor Drain Tank Room 7126 in the radwaste building to support a task performed by two radwaste operators. The health physics technician surveyed the immediate work area and informed the operators that general work area dose rates were 10 millirem per hour. Based on this information, operators entered the posted high radiation area on a radiation work permit that had an electronic dosimeter dose rate set point of 50 millirem per hour. One of the operators received a dose rate alarm while performing the task, the operators exited the area, and contacted health physics personnel. Subsequent investigation identified that a comprehensive survey of the entire room was not performed. Follow-up surveys indicated that dose rates in the room were as high as 150 millirem per hour at 30 centimeters from the floor drain tank. This issue has been entered into the licensees corrective action program.

The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Exposure Control, and affected the cornerstone objective because workers could have received additional radiation dose. The finding was processed through the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance because it was not an as low as is reasonably achievable finding, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.

Additionally, this finding has a crosscutting aspect in the area of human performance related to work controls because the failure to incorporate job site conditions impacted the margin of radiological safety provided by an adequate survey. (Section 2OS1)

Licensee-Identified Violations

None

REPORT DETAILS

Summary of Plant Status

Wolf Creek operated at full power for the inspection period.

REACTOR SAFETY

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

1R01 Adverse Weather Protection

.1 Readiness for Seasonal Susceptibilities

a. Inspection Scope

The inspectors completed a review of the licensee's readiness of seasonal susceptibilities involving winter weather. The inspectors:

(1) reviewed plant procedures, the Updated Safety Analysis Report (USAR), and Technical Specifications (TSs) to ensure that operator actions defined in adverse weather procedures maintained the readiness of essential systems;

(2) walked down portions of the system listed below to ensure that adverse weather protection features were sufficient to support operability including the ability to perform safe shutdown functions;

(3) evaluated operator staffing levels to ensure the licensee would maintain the readiness of essential systems required by plant procedures; and

(4) reviewed the corrective action program to determine if the licensee identified and corrected problems related to adverse weather conditions.

C January 16, 2007, Licensee verification for cold weather preparations of essential service water warming line flow Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.2 Readiness For Impending Adverse Weather Conditions

a. Inspection Scope

On January 12, 2007, the inspectors completed a review of the licensee's readiness for impending adverse weather, specifically with respect to freezing rain and severe cold weather. The inspectors:

(1) reviewed plant procedures, the USAR, and TSs to ensure that operator actions defined in adverse weather procedures maintained the readiness of essential systems;

(2) walked down portions of the facility grounds to identify hazards that could potentially be blown by heavy winds; and

(3) reviewed plant modifications, procedure revisions, and operator work arounds to determine if recent facility changes challenged plant operation.

• January 12, 2007, Licensee preparations for freezing rain and severe cold weather Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Partial System Walkdowns

a. Inspection Scope

The inspectors:

(1) walked down portions of the risk important systems listed below and reviewed plant procedures and documents to verify that critical portions of the selected systems were correctly aligned and

(2) compared deficiencies identified during the walkdown to the licensee's USAR and corrective action program to ensure problems were being identified and corrected.

• January 10, 2007, Residual heat removal relief valve header draining for planned maintenance

• January 17, 2007, Residual heat removal Train A while residual heat removal Train B was inoperable for planned maintenance

• January 18, 2007, Emergency Diesel Generator A while Emergency Diesel Generator B was inoperable during testing

• January 24, 2007, Motor-driven auxiliary feedwater pump Train B

• January 30, 2007, Emergency fuel oil system lineup, during Emergency Diesel Generator A run Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed five samples.

b. Findings

No findings of significance were identified

.2 Complete Walkdown

The inspectors:

(1) reviewed plant procedures, drawings, the USAR, TSs, 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.

C January 16, 2007, Essential service water Train A Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified

1R05 Fire Protection

.1 Fire Protection Tours

a. Inspection Scope

The inspectors walked down the plant areas listed below to assess the material condition of active and passive fire protection features, their operational lineup, and their operational effectiveness. The inspectors:

(1) verified that transient combustibles and hot work activities were controlled in accordance with plant procedures;

(2) observed the condition of fire detection devices to verify they remained functional;

(3) observed fire suppression systems to verify they remained functional;

(4) verified that fire extinguishers and hose stations were provided at their designated locations and that they were in a satisfactory condition;

(5) verified that passive fire protection features (electrical raceway barriers, fire doors, fire dampers, steel fire proofing, penetration seals, and oil collection systems) were in a satisfactory material condition;

(6) verified that adequate compensatory measures were established for degraded or inoperable fire protection features; and

(7) reviewed the corrective action program to determine if the licensee identified and corrected fire protection problems.

C January 8, 2007, Turbine building 2000' C

January 10, 2007, Containment 2000' C

January 11, 2007, Turbine-driven auxiliary feedwater pump room C

January 16, 2007, Essential service water pump house C

January 19, 2007, Emergency Diesel Generator A C

January 25, 2007, Turbine building 2033' C

February 26, 2007, Lower cable spreading room C

February 28, 2007, Auxiliary building 1974' Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed eight samples.

b. Findings

Introduction:

A self-revealing Green noncited violation (NCV) of TS 5.4.1.d was identified for failure to implement fire protection impairment control permit requirements and compensatory measures when operators incorrectly disabled three fire detectors in the auxiliary building.

Description:

In January 2007, a modification requiring welding and grinding was being performed in the radwaste building. This work required a transient ignition source permit and a fire protection impairment control permit disabling five fire detectors in the area of the hot work to prevent inadvertent alarms.

On January 23, 2007, following restoration from work, it was discovered that three fire detectors in the auxiliary building 1974' level were disabled instead of fire detectors in the radwaste building as stated on the fire protection impairment control permit. The detectors in the auxiliary building were disabled without a proper fire impairment control permit and the required compensatory roving hourly fire watch for a period of approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> as required by Administrative Procedure AP 10-103, Fire Protection Impairment Control, Revision 21. During interviews with the control room operator that disabled the points, the inspectors noted that the operator normally requested peer checks for verification of disabled points but due to high workload, multiple distractions, and perceived time pressure, the operator failed to utilize any human error prevention tools.

Analysis:

The failure to implement fire protection impairment control permit requirements and establish compensatory measures for the auxiliary building 1974' level was considered a performance deficiency. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRCs regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this issue relates to the fire example of protection against external factors attribute because the operator removed the detectors from service without preplanning or ensuring compensatory measures where in place.

The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process. The inspectors determined that the finding was of very low safety significance because it involved compensatory measures for the fixed fire protection system and was assigned a low degradation rating since less than 10 percent of the fire detectors in the area were disabled. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to apply appropriate human error preventions techniques such as self and peer-checking prior to removing the detectors from service.

Enforcement:

TS 5.4.1.d requires that written procedures be established, implemented, and maintained covering activities related to fire protection program implementation.

Administrative Procedure AP 10-103, Fire Protection Impairment Control, Revision 21, requires, in part, fire protection impairment control permit shall be prepared in order to determine the appropriate compensatory measures and track the impairment. Contrary to the above, on January 23, 2007, three fire detectors were disabled in the auxiliary building 1974' without implementing a fire protection impairment control permit and establishing compensatory measures because the operator did not correctly verify that the applicable fire detectors were removed from service before proceeding with the maintenance. This issue and the corrective actions are being tracked by the licensee in Condition Report 2007-000298. Because the finding is of very low safety significance and has been entered into the corrective action program, this violation is being treated as NCV 05000482/2007002-01, failure to implement fire protection impairment control permit requirements and compensatory measures.

.2 Annual Fire Drill

a. Inspection Scope

The inspectors observed a fire brigade drill to evaluate the readiness of licensee personnel to prevent and fight fires, including the following aspects:

(1) the number of personnel assigned to the fire brigade,

(2) use of protective clothing,

(3) use of breathing apparatuses,

(4) use of fire procedures and declarations of emergency action levels,

(5) command of the fire brigade,

(6) implementation of prefire strategies and briefs,

(7) access routes to the fire and the timeliness of the fire brigade response,

(8) establishment of communications,

(9) effectiveness of radio communications,

(10) placement and use of fire hoses,

(11) entry into the fire area,

(12) use of fire fighting equipment,

(13) searches for fire victims and fire propagation,

(14) smoke removal,

(15) use of prefire plans,

(16) adherence to the drill scenario,

(17) performance of the post-drill critique, and

(18) restoration from the fire drill. The licensee simulated a fire in the area listed below.

C January 25, 2007, Turbine building 2033' Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

Semiannual Internal Flooding

a. Inspection Scope

For the area listed below, the inspectors:

(1) reviewed the USAR, the flooding analysis, and plant procedures to assess seasonal susceptibilities involving internal flooding;

(2) reviewed the corrective action program to determine if the licensee identified and corrected flooding problems;

(3) inspected underground bunkers/manholes to verify the adequacy of

(a) sump pumps,

(b) level alarm circuits,

(c) cable splices subject to submergence, and

(d) drainage for bunkers/manholes;

(4) verified that operator actions for coping with flooding can reasonably achieve the desired outcomes; and

(5) walked down the areas listed below to verify the adequacy of

(a) equipment seals located below the flood line,

(b) floor and wall penetration seals,

(c) watertight door seals,

(d) common drain lines and sumps,

(e) sump pumps, level alarms, and control circuits, and

(f) temporary or removable flood barriers.

C Motor-driven auxiliary feedwater pump Room A Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

a. 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 licensees heat exchanger inspections adequately assessed the state of cleanliness of their tubes, and

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

C February 5, 2007, Essential service water heat Exchanger A Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

Resident Inspector Quarterly Review

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 is listed below.

C March 8, 2007, Loss of instrument air due to tampering Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

Resident Inspector Quarterly Review

a. Inspection Scope

The inspectors reviewed the maintenance activity listed below to:

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

(2) verify the appropriate handling of degraded SSC functional performance;

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

(4) evaluate the handling of SSC issues reviewed under the requirements of the maintenance rule, 10 CFR Part 50, Appendix B, and TSs.

• February 14, 2007, KC HV-0253 valve failure Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

Risk Assessment and Management of Risk The inspectors reviewed the assessment activities listed below to verify:

(1) performance of risk assessments when required by 10 CFR 50.65 (a)(4) and licensee procedures prior to changes in plant configuration for maintenance activities and plant operations;

(2) the accuracy, adequacy, and completeness of the information considered in the risk assessment;

(3) that the licensee recognizes, and/or enters as applicable, the appropriate licensee-established risk category according to the risk assessment results and licensee procedures; and

(4) that the licensee identified and corrected problems related to maintenance risk assessments.

• January 22, 2007, Corrected daily risk assessment for switchyard work

• March 3, 2007, Corrected daily risk assessment for service air Compressor A work

• March 14, 2007, Turbine-driven auxiliary feedwater pump planned outage Emergent Work Control 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 USAR and corrective action program to determine if the licensee identified and corrected risk assessment and emergent work control problems.

• January 3, 2007, Emergent work on Switchyard Disconnect 345-53

• February 6, 2007, Emergent work on essential service water Pump A fan damper

• February 9, 2007, Emergent work on normal service water Valve 1WS0002A Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed six samples.

b. Findings

During an attempt to open service water pump discharge Valve 1WS0002A on February 8, 2007, a plant operator initially attempted to open the valve with the pump secured but was unsuccessful because the valves manual handwheel was too difficult to turn against the system back pressure. However, Procedure SYS EA-120, Service Water System Startup Revision 40A, states that service water Pump 1WS01PA be started prior to opening the discharge valve to equalize system back pressure. A second attempt to open the valve was made with the service water pump running to decrease differential pressure across the valve. The second attempt was also unsuccessful as the valves handwheel was still too difficult to operate. During the second attempt, the operator placed a wrench across the valvess handwheel to increase leverage and open the valve. The second attempt was not successful in opening the valve and resulted in damage to the handwheel and valve actuator internals due to the excessive torsion caused by using a wrench. Wolf Creek Standing Order 1, Valve Setup and Operation, Revision 37 as well as Enertech (the valve vendor)

Procedure PA 97283, Installation, Operation and Maintenance Instructions - Enertech Permaseat Valves, MAK, prohibit the use of a wrench and use of excessive force on the handwheel. Subsequently, the valves restoration position on the tracking document was then changed to closed. The valve was repaired as emergent work and returned to service after approximately 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> of being inoperable. The inspectors questioned whether the plant operator used appropriate guidance which resulted in applying excessive torque with a wrench on the valves handwheel. This issue is captured in the licensees corrective action program under Condition Report 2007-000543. Because additional followup is needed to determine the significance, this issue is being treated as an Unresolved Item URI 05000482/2007002-02, use of wrench on manual service water valve handwheel.

1R15 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 TSs;

(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.

• January 18, 2007, Main steam isolation valve packing leak

• January 24, 2007, Safety-related load center rear panel bolt qualification

• February 1, 2007, Component cooling water Pump B room cooler airflow lower than rated

• March 16, 2007, Component cooling water Pump A room cooler motor wiring Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed four samples.

b. Findings

No findings of significance were identified.

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.

• January 17, 2007, Residual heat removal Train B inservice pump test following maintenance

• January 22, 2007, Loop 1 main steam isolation Valve HV14 yellow accumulator air regulator test following repair

• January 31, 2007, Containment cooling fan Train A 480V ESF circuit breaker testing following replacement

• February 8, 2007, Main turbine load limit potentiometer following repair

• February 15, 2007, Emergency Diesel Generator B fast start following maintenance

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed five samples.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the USAR, procedure requirements, and TSs to ensure that the listed surveillance activities demonstrated that the SSCs tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the following significant surveillance test attributes were adequate:

(1) preconditioning;

(2) evaluation of testing impact on the plant;

(3) acceptance criteria;

(4) test equipment;

(5) procedures;

(6) jumper/lifted lead controls;

(7) test data;

(8) testing frequency and method demonstrated TS operability;

(9) test equipment removal;

(10) restoration of plant systems;

(11) fulfillment of American Society of Mechanical Engineers code requirements;

(12) updating of performance indicator (PI) data;

(13) engineering evaluations, root causes, and bases for returning tested SSCs not meeting the test acceptance criteria were correct;

(14) reference setting data; and

(15) annunciators and alarms setpoints. The inspectors also verified that the licensee identified and implemented any needed corrective actions associated with the surveillance testing:

• January 11, 2007, Impulse chamber pressure transmitter surveillance

• January 18, 2007 Hot restart of Emergency EDG NE02

• January 24, 2007, Turbine-driven auxiliary feedwater pump test

• January 28, 2007, RHR system inservice Pump B test

• February 6, 2007, Incore-excore detector axial flux calibration

• February 15, 2007, Spent fuel pool gate operations Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed six samples.

b. Findings

No findings of significance was identified.

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the USAR, plant drawings, procedure requirements, and TSs to ensure that the temporary modification(s) listed below was properly implemented.

The inspectors:

(1) verified that the modification did not have an affect on system operability/availability;

(2) verified that the installation was consistent with modification documents;

(3) ensured that the post-installation test results were satisfactory and that the impact of the temporary modifications on permanently installed SSCs were supported by the test;

(4) verified that the modifications were identified on control room drawings and that appropriate identification tags were placed on the affected drawings; and

(5) verified that appropriate safety evaluations were completed. The inspectors verified that licensee identified and implemented any needed corrective actions associated with temporary modifications.

• January 9, 2007, AEFCV-520 feedwater regulating valve furmanite injection Documents reviewed by the inspectors are listed in the attachment.

The inspector completed one sample.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

For the below listed drill and simulator-based training evolution contributing to drill/exercise performance and emergency response organization PIs, 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 Nuclear Energy Institute (NEI) 99-02 documents acceptance criteria.

• February 22, 2007, Reactor coolant system leak with loss of reactor coolant pump Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access Control To Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess licensee performance in implementing physical and administrative controls for airborne radioactivity areas, radiation areas, high radiation areas, and worker adherence to these controls. The inspector used the requirements in 10 CFR Part 20, the TSs, and licensees procedures required by TS as criteria for determining compliance. During the inspection, the inspector interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspector performed independent radiation dose rate measurements and reviewed the following items:

• PI events and associated documentation packages reported by the licensee in the Occupational Radiation Safety Cornerstone

• Controls (surveys, posting, and barricades) of three radiation, high radiation, or airborne radioactivity areas

• Radiation work permits, procedures, engineering controls, and air sampler locations

• Conformity of electronic personal dosimeter alarm set points with survey indications and plant policy; workers knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarms.

• Physical and programmatic controls for highly activated or contaminated materials (nonfuel) stored within spent fuel and other storage pools.

• Self-assessments, audits, licensee event reports (LERs), and special reports related to the access control program since the last inspection

• Corrective action documents related to access controls

• Licensee actions in cases of repetitive deficiencies or significant individual deficiencies

• Radiation work permit briefings and worker instructions

• Adequacy of radiological controls such as, required surveys, radiation protection job coverage, and contamination controls during job performance

• Changes in licensee procedural controls of high dose rate - high radiation areas and very high radiation areas

• Controls for special areas that have the potential to become very high radiation areas during certain plant operations

• Posting and locking of entrances to all accessible high dose rate - high radiation areas and very high radiation areas

• Radiation worker and radiation protection technician performance with respect to radiation protection work requirements The inspector completed 18 of the required 21 samples.

b. Findings

Introduction.

The inspector reviewed a self-revealing Green NCV of 10 CFR 20.1501(a)because the licensee failed to perform an adequate survey in a high radiation area. The violation had very low safety significance.

Description.

On March 7, 2007, two operators and a health physics technician were preparing to commence work in the Floor Drain B Tank Room 7126 in the radwaste building. That room was posted a high radiation, contaminated area. The task was to drain the liquid radwaste system by running a tube from the system vent to the floor drain in the room. Prior to the work beginning, a health physics technician performed a survey in the work area and determined that the highest dose rate in the area was 30 mrem/hr near the floor drain between two 55 gallon drums of oil socks and that the general area dose rates in the immediate work area were 10 millirem per hour. Based on this information the radwaste operators logged in on Radiation Work Permit 070065, Task 2, which had electronic dosimeter alarm set points of 5 mrem and 50 mrem/hr. Based on the survey, these set points should have been sufficient to complete the task. However, during the performance of the task, one of the operators received an electronic dosimeter dose rate alarm as the operator bent over to adjust the tubing and place it in the floor drain. The licensees investigation of this event determined that a complete and thorough survey of the room was not performed and the health physics technician did not provide direct coverage at the time of the alarm. Additional surveys by the health physics technician and other health physics staff determined that the dose rates were as high as 150 mrem per hour at 30 centimeters from the floor drain tank and as much as 75 millirem per hour 30 centimeters from one of the two drums by the floor drain.

Analysis.

The failure to perform an adequate survey was a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Exposure Control, and affected the cornerstone objective to ensure the adequate protection of a workers health and safety from exposure to radiation because workers could have received additional radiation dose.

Because the finding involved the potential for unplanned, unintended dose resulting from conditions that were contrary to NRC regulations, the finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The finding was determined to be of very low safety significance because:

(1) it was not an as low as is reasonably achievable (ALARA) finding,

(2) there was no personnel overexposure, (3)there was no substantial potential for personnel overexposure, and

(4) the finding did not compromise licensees ability to assess dose. Additionally, this finding has a crosscutting aspect in the area of human performance related to work controls because the failure to incorporate job site conditions impacted the margin of radiological safety provided by an adequate survey.

Enforcement.

Part 20.1501(a) of Title 10 of the Code of Federal Regulations requires that each licensee make or cause to be made surveys that may be necessary to comply with the regulations in Part 20 to determine the extent and magnitude of radiation levels and to evaluate the radiological hazards. Pursuant to 10 CFR 20.1003, survey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation. Contrary to 10 CFR 20.1501(a), on March 7, 2007, the licensee failed to perform an adequate survey of Floor Drain Tank Room 7126 of the radwaste building to assure compliance with 10 CFR 20.1201(a), which limits radiation exposure to occupational workers to 5.0 rem total effective dose equivalent. This violation was entered into the licensees corrective action program as Condition Report 2007-000941.

Because this finding is of very low safety significance and was entered into the licensees corrective action program, it is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000482/2007002-03, failure to perform an adequate survey in a high radiation area.

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual and collective radiation exposures ALARA. The inspector used the requirements in 10 CFR Part 20 and the licensees procedures required by TSs as criteria for determining compliance. The inspector interviewed licensee personnel and reviewed:

• Use of engineering controls to achieve dose reductions and dose reduction benefits afforded by shielding

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

• Source-term control strategy or justifications for not pursuing such exposure reduction initiatives

• Specific sources identified by the licensee for exposure reduction actions, priorities established for these actions, and results achieved since the last refueling cycle

• Effectiveness of self-assessment activities with respect to identifying and addressing repetitive deficiencies or significant individual deficiencies The inspector completed 3 of the required 15 samples and 2 of the optional samples.

b. Findings

No findings of significance were identified.

4OA1 PI Verification

Cornerstone: Initiating Events

The inspectors sampled licensee submittals for the PIs listed below. The definitions and guidance of Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Indicator Guideline, Revision 4, were used to verify the licensees basis for reporting each data element in order to verify the accuracy of PI data reported during the assessment period.

The inspectors reviewed LERs, monthly operating reports, and operating logs as part of the assessment. Licensee PI data were also reviewed against the requirements of Procedure AP 26A-007, "NRC Performance Indicators," Revision 4, and "Performance Improvement and Learning Desktop Instruction, NRC Performance Indicator Program Owner Guidance, Revision 2. The inspectors reviewed various licensee indicator input information to determine the accuracy and completeness of the PI.

C Unplanned scrams per 7,000 critical hours C

Unplanned scrams with loss of normal heat removal C

Unplanned power changes per 7,000 critical hours The inspectors completed three samples in this cornerstone.

Cornerstone: Barrier Integrity

The inspectors sampled licensee submittals for the two PIs listed below for the period, September 30, 2005, through January 1, 2007. The definitions and guidance of NEI 99-02, Regulatory Assessment Indicator Guideline, Revision 4, were used to verify the licensees basis for reporting each data element in order to verify the accuracy of PI data reported during the assessment period. The inspectors:

(1) reviewed reactor coolant system chemistry sample analyses for dose equivalent Iodine-131 and compared the results to the TS limit;

(2) observed a chemistry technician obtain and analyze a reactor coolant system sample;

(3) reviewed operating logs and surveillance results for measurements of reactor coolant system identified leakage; and

(4) observed a surveillance test that determined reactor coolant system identified leakage.

C Reactor coolant system specific activity C

Reactor coolant system leakage The inspectors completed two samples in this cornerstone.

Occupational Radiation Safety Cornerstone Occupational Exposure Control Effectiveness The inspector reviewed licensee documents from October 1 through December 31, 2006.

The review included corrective action documentation that identified occurrences in locked high radiation areas (as defined in the licensees TSs), very high radiation areas (as defined in 10 CFR 20.1003), and unplanned personnel exposures (as defined in NEI 99-02, "Regulatory Assessment Indicator Guideline," Revision 4). Additional records reviewed included ALARA records and whole body counts of selected individual exposures. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the PI data. In addition, the inspector toured plant areas to verify that high radiation, locked high radiation, and very high radiation areas were properly controlled. PI definitions and guidance contained in NEI 99-02, Revision 4, were used to verify the basis in reporting for each data element.

The inspector completed the required one sample in this cornerstone.

Public Radiation Safety Cornerstone Radiological Effluent Technical Specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences The inspector reviewed licensee documents from October 1 through December 31, 2006.

Licensee records reviewed included corrective action documentation that identified occurrences for liquid or gaseous effluent releases that exceeded PI thresholds and those reported to the NRC. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the PI data. Performance Indicator definitions and guidance contained in NEI 99-02, Revision 4, were used to verify the basis in reporting for each data element.

The inspector completed the required one sample in this cornerstone.

b. Findings

No findings of significance were identified

4OA2 Identification and Resolution of Problems

.1 Routine Review of Identification and Resolutions of Problems

a. Inspection Scope

The inspectors performed a daily screening of items entered into the licensee's corrective action program. This assessment was accomplished by reviewing work requests, work orders, and performance improvement requests, and attending corrective action review and work control meetings. The inspectors:

(1) verified that equipment, human performance, and program issues were being identified by the licensee at an appropriate threshold and that the issues were entered into the corrective action program;

(2) verified that corrective actions were commensurate with the significance of the issue; and

(3) identified conditions that might warrant additional followup through other baseline inspection procedures.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.2 Selected Issue Follow-up Inspection

a. Inspection Scope

In addition to the routine review, the inspectors selected Condition Report 2006-003768 below listed issues 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.

C CR 2006-003768, SJ HV-137 found in unanticipated position Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified

.3 Cumulative Review of the Effects of Operator Workarounds

a. Inspection Scope

The inspectors reviewed the cumulative effects of operator workarounds to determine:

(1) the reliability, availability, and potential for misoperation of a system;

(2) if multiple mitigating systems could be affected;

(3) the ability of operators to respond in a correct and timely manner to plant transients and accidents; and

(4) if the licensee has identified and implemented appropriate corrective actions associated with operator workarounds.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.4 Radiation Protection

a. Inspection Scope

The inspector evaluated the effectiveness of licensees problem identification and resolution process with respect to the following inspection areas:

• Access Control to Radiologically Significant Areas (Section 2OS1)

• ALARA Planning and Controls (Section 2OS2)

b. Findings and Observations

No findings of significance were identified. However, during a review of corrective action documents, the inspector identified that the licensee did not fully resolve and document the actions taken for at least three examples involving personnel failing to obtain an exit whole body count upon termination from the site. The licensee closed the corrective action document based on actions taken, but did not document that the actions included completing an evaluation of each individuals potential internal dose as required by procedure. Additionally, the licensee did not fully evaluate why the whole body count was not performed. The inspector noted that the licensee was issued an NCV in Inspection Report 05000482/2005004 for failing to perform an evaluation of internal dose and document their actions. The licensees corrective action in response to the violation was to perform the internal dose evaluation and complete the appropriate documentation.

However, one of the lessons learned from the 2005 violation was that a contributing factor that led to the performance deficiency was a failure to document the decisions for not performing an exit whole body count. Again, during this inspection, the inspector observed that documentation of actions taken to close corrective action documents associated with not performing an exit whole body count were not complete.

The inspector also reviewed two corrective action documents associated with individuals who alarmed the portal radiation monitors at the protected area egress point and failed to notify health physics prior to departing the protected area. One of the individuals received a medical uptake, but was not on the approved list for egress due to medical uptakes because the individual had not informed the radiation protection manager about the medical procedure. On another occasion, an individual was attempting to remove a piece of stoneware from the site and alarmed the monitors. The inspector observed that the licensees corrective actions for these two events were limited in scope and verified by the inspector to be incomplete or ineffective. The corrective actions were limited in scope because they only targeted the groups specifically involved in these events, namely security and supply chain services. The inspector determined that corrective actions were incomplete or ineffective because the inspector interviewed security officers who were not provided training or information about the events. The inspector also noted that licensees procedures required that the radiation protection manager be informed about medical uptakes of radioisotopes, however, this procedure is only applicable to radiation workers, and thus, nonradiation workers were not informed about this requirement. Furthermore, this requirement is not presented to all employees as part of the licensees general employee training program.

The licensee acknowledged the inspectors observations and has modified their problem identification and resolution program by requiring that all corrective actions associated with radiological issues be reviewed by the health physics department prior to closure.

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

.1 Unexpected Loss of Greater than 75 Percent of Control Room Annunciators on

January 26, 2007

a. Inspection Scope

The inspectors:

(1) reviewed operator logs, plant computer data, and/or strip charts for the above listed event to evaluate operator performance in coping with nonroutine events and transients;

(2) verified that operator actions were in accordance with the response required by plant procedures and training; and

(3) verified that the licensee has identified and implemented appropriate corrective actions associated with personnel performance problems that occurred during the events sampled.

b. Findings

(1)

Introduction:

A Green NCV of TS 5.4.1.a was identified by the inspectors for failure to maintain sufficient records (logs) to furnish evidence of events significant to plant safety.

Description:

On January 26, 2007, electrical maintenance commenced a scheduled replacement of main control board (MCB) Annunciator Power Supply E1PS5. During the power supply replacement, a loss of 8.7 percent of the annunciators was expected.

However, during de-termination of the power supply leads, an unexpected loss of a significant number of the MCB annunciators occurred. Since the loss of MCB annunciators occurred near the end of the shift, the oncoming shift manager was present in the control room. Based on discussions between the off-going shift manager and the oncoming shift manger, they agreed that greater than 75 percent of the MCB annunciators were lost. Subsequently, due to the large number of annunciator inputs that were lost, the plant computer became overloaded and stopped updating. Based on these indications, the control room operators would need to evaluate emergency action levels and TS requirements. Emergency Action Level-6-LEP/AC8 required an ALERT declaration if greater than 75 percent of MCB annunciators were lost for greater than 15 minutes.

The inspectors reviewed log entries made approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the event and noted that the operator logs did not contain any information to determine the start or end of the unexpected loss of annunciators. The inspectors discovered during interviews with the operations crew that was on watch during the event, that no information was recorded or kept during the event. Administrative Procedure AP 21-001, Conduct of Operations, Revision 36A, requires operators to make plant log entries of potentially reportable occurrences, entries that could be useful in reconstructing events, and events significant to plant safety. However, the logs were not updated until several hours later based on verbal accounts provided to the oncoming crew. The after the fact log entries were entered at the original declared start and end times by the oncoming shift manager with no indication of a late entry noted in the logs. The inspectors noted that the after the fact log entries still provided insufficient data to reconstruct the activities related to the loss of annunciators. The information required to be contained in operator logs is considered to be important because it provides quality evidence of activities associated with determining emergency action level declarations and meeting TS action statements for events.

Analysis:

The failure to adequately document times and information for the loss of annunciators was considered to be a performance deficiency. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRCs regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. This finding was more than minor because it could impact the operators ability to accurately implement emergency action levels and TS action statements and if left uncorrected, this type of insufficient documentation could become a more significant safety concern.

The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609, Appendix M, "Significance Determination Process using Qualitative Criteria," and determined that the finding was of very low safety significance based on the fact that the loss of annunciators challenged the emergency action level time requirements but was restored prior to exceeding emergency action level and TS action time requirements. This observation was based on the inspectors review of video recordings and interviews to bound the times for the beginning and end of the event. The NRC management review concurred with the determination of very low safety significance. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively communicate expectations regarding plant operating log entries in accordance with procedural requirements.

Enforcement:

Technical Specification 5.4.1.a requires that written procedures be established, implemented, and maintained covering activities related to procedures recommended in Regulatory Guide 1.33 Revision 2, Appendix A, 1978. Regulatory Guide 1.33, Section 1(h), Administrative Procedures, requires log entries. Administrative Procedure AP 21-001, Conduct of Operations, Revision 36A, requires, in part, plant log entries of potentially reportable occurrences, entries that could be useful in reconstructing events, and events significant to plant safety. Contrary to the above, logkeeping and records were not sufficiently maintained to furnish evidence to reconstruct activities related to the loss of annunciators on January 26, 2007. This issue and the corrective actions are being tracked by the licensee in Condition Report 2007-000362. Because the finding is of very low safety significance and has been entered into the corrective action program, this violation is being treated as NCV 05000482/2007002-04, failure to maintain sufficient records to furnish evidence of events significant to plant safety.

(2)

Introduction:

A Green self-revealing finding was identified regarding inadequate engineering drawings used as guidance to replace MCB annunciator power supplies resulting in a loss of all MCB annunciators.

Description:

On January 26, 2007, electrical maintenance commenced a scheduled replacement of MCB Annunciator Power Supply E1PS5. During the power supply replacement, a loss of 8.7 percent of the annunciators was expected. However, during de-termination of the power supply leads, an unexpected loss of a significant number of the MCB annunciators occurred.

During the planning review of Work Order 06-280217-003, Replace power supply RK045E1PS5, the electricians brought forth a concern about the daisy chaining of the leads associated with the MCB power supplies and not knowing what effect removing a power supply would have on additional MCB annunciators. System engineering reviewed vendor drawings and determined that only the expected annunciators would be lost. The vendor drawings only consisted of discrete wire connections from the power supply to the logic bus and did not show interconnections with any other power supplies. Although, it was acknowledged by system engineering that there were numerous daisy chained connections not shown on the vendor drawings, no further reviews or research was conducted.

During the licensees root cause investigation, it was discovered that 100 percent of the MCB annunciators became inoperable during the power supply replacement. After tracing the wiring between power supplies, it was concluded that lifting a single wire on the E1PS5 power supply disconnected all the annunciator cards from their respective power supplies. The root cause analysis determined that the vendor drawings did not show the interconnecting wiring identifying point to point connections associated with the MCB power supplies.

Analysis:

The failure to maintain drawings technically accurate and reflect the as-built condition of the plant was considered to be a performance deficiency. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRCs regulatory function, and the finding was not the result of any willful violation of NRC requirements or Wolf Creek procedures. The inspectors determined that the finding was more than minor because it impacted the maintenance technicians ability to accurately plan and implement work, resulting in the annunciator system being adversely affected and could be reasonably viewed as a precursor to a significant event.

The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609, Appendix M, "Significance Determination Process using Qualitative Criteria," and determined that the finding was of very low safety significance because the finding did not result in a loss of a system safety function or a loss of risk significant equipment for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The NRC management review concurred with the determination of very low safety significance. This finding has a crosscutting aspect in the human performance area associated with the resources component because the licensee failed to maintain complete, accurate and up-to-date design documentation.

Enforcement:

Enforcement action does not apply because the performance deficiency did not involve a violation of a regulatory requirement. This issue and the corrective actions are being tracked by the licensee in Condition Report 2007-000362. Because the finding is of very low safety significance and has been entered into the corrective action program, this finding is being treated as a Finding: FIN05000482/2007002-05, failure to maintain drawings technically accurate.

.2 Reactor Overpower During Neutron Detector Calibration on February 6, 2007

a. Inspection Scope

The inspectors:

(1) reviewed operator logs, plant computer data, and/or strip charts for the above listed evolution to evaluate operator performance;

(2) reviewed operator actions with the response required by plant procedures and training;

(3) verified that the licensee has identified and implemented appropriate corrective actions associated with personnel performance problems that occurred during the event sampled; and

(4) interviewed plant personnel.

b. Findings

Introduction:

A Green NCV of TS 5.4.1.a occurred because operators did not take timely action to lower power below the licensed thermal limit of 3565 Mwt in accordance with GEN 00-004, Power Operation.

Description:

During a control room walk down on February 6, 2007, the inspectors identified an overpower condition. At the time, an incore to excore neutron detector calibration was in progress. The inspectors questioned the control room supervisor and he was aware of the trend in reactor power and he stated that if the instantaneous reactor power (i.e., the plant computer calorimetric that runs every 10 to 30 seconds)exceeded 3575 MWt he would order action to reduce power. The power level was maintained and the operating crew did not insert negative reactivity until after the neutron detector calibration was complete. During this evolution, the reactor exceeded licensed thermal power of 3565 MWt for approximately 58 minutes, peaking at 3566.5 MWt according to the plant computers 10-minute calorimetric. After the neutron detector calibration was completed, operators added boron to the reactor coolant system to reduce power below 100 percent.

The inspectors reviewed recorded plant parameter data, operator actions, and the plant operating procedures used during the evolution. During the inspection, the licensee cited Procedure GEN 00-004, Power Operation, Attachment B, Step B.1.1 which states, in part, that exceeding 3565 MWt is permitted only as a result of transients or computer point fluctuations. The inspectors judged that allowing reactor power to ascend above 100 percent for nearly an hour was not a transient; however, computer point fluctuations on the plant computer calorimetric that calculates every 10 to 30 seconds can fluctuate above and below 100 percent with each iteration of the calculation. The inspectors judged the latter to be appropriate but not applicable in this case because the 10-minute power average and the 1-hour power average trended above 100 percent power due to prior dilutions and not fluctuations. The inspectors also found that Procedure GEN 00-004, Attachment B, Step B.1.1 states, in part,

(1) that the 10 minute average should not exceed 3565 MWt; and

(2) that action shall be taken to reduce power if the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average exceeds 3565 MWt to prevent the 8-hour power average from exceeding 3565 MWt. However, operators did not initiate action in accordance with Step B.1.1 when the 10 minute average exceeded 3565 MWt until approximately 40 minutes had elapsed. The inspectors reasoned that it would have required extended operation above 3565 MWt to increase the 8-hour power average above 3565 MWt.

Lastly, the inspectors could not locate any license condition or regulation that permits operation above the licensed power level for any period of time as articulated in GEN 00-004.

Analysis:

The inspectors determined that the failure to maintain steady state reactor power at or below the licensed thermal power limit was a performance deficiency.

Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRCs 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 is associated with the configuration control attribute for the Barrier Integrity Cornerstone; and, it affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radio nuclide releases caused by accidents or events. Specifically, this issue relates to the reactor manipulation example of the configuration control attribute because operating the reactor above licensed power for an extended period can challenge fuel cladding integrity during events by reducing calculated margins to unacceptable fuel cladding temperatures.

The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because the fuel cladding barrier was affected but did not affect the reactor coolant system or containment barriers. Further, the actual increase in reactor power of less than two percent did not result in a challenge to design limits or other overpower protection features. As such, the assumptions contained in the safety analysis remained valid for this event. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee did not ensure that licensed operators used conservative assumptions in their decision making when reactor power increased above the licensed limit for an extended period.

Enforcement:

Technical Specification 5.4.1.a, requires the procedures in Regulatory Guide 1.33, Revision 2, Appendix A to be established, implemented, and maintained.

Regulatory Guide 1.33 Appendix A, Part 2.g requires procedures for Power Operation and Process Monitoring, and License Condition 2.C.(1) requires reactor power to be maintained at or below 3565 MWt. The TS required procedure and license condition limit of 3565 MWt are implemented by Procedure GEN 00-004, Power Operation, Revision 55. Contrary to the above, on February 6, 2007, Wolf Creek power exceeded 3565 MWt for approximately 58 minutes because operators believed that Procedure GEN 00-004, Attachment B, Step B.1.1 permitted operation above the licensed power limit. This issue and the corrective actions are being tracked by the licensee in Condition Report 2007-001352. Because the violation was of very low safety significance and the issue was captured in the licensees corrective action program, this violation is being treated as an NCV consistent with Section VI.A of the NRC Enforcement Policy, NCV 05000482/2007002-06, failure to maintain steady state reactor power at or below the licensed thermal power limit.

.3 Removal of Both Containment Purge Radiation Monitors GTRE22 and GTRE33 From

Service

a. Inspection Scope

The inspectors:

(1) reviewed operator logs, plant computer data, and/or strip charts for the above listed evolution to evaluate operator performance in coping with nonroutine events and transients;

(2) reviewed operator actions with the response required by plant procedures and training;

(3) verified that the licensee has identified and implemented appropriate corrective actions associated with personnel performance problems that occurred during the event sampled; and

(4) interviewed plant personnel.

b. Findings

Introduction:

A self-revealing Green NCV of TS 5.4.1.a occurred on February 20, 2007, when a chemistry technician inadvertently removed both containment purge radiation Monitors GTRE22 and GTRE33 from service at the same time.

Description:

During planned maintenance on the safety-related GTRE33 containment purge radiation monitor, a chemistry technician inadvertently removed the incorrect containment purge radiation monitor from service. GTRE33 was removed from service (i.e. the input signal to the engineered safety features actuation system was bypassed)so that Procedure STN SP-133, could be performed to calibrate the device, and Procedure CHS AX-G02, Exchange of Radioactive Gaseous Monitor Particulate and Iodine Filters, could be performed to install a new particulate filter and a new iodine cartridge. As part of the evolution, the instrumentation and controls department requested the shift chemist to perform the filter and iodine cartridge changes for GTRE33. After contacting the control room, the shift chemist went to GTRE22 and incorrectly removed the radiation monitor from service. Instrumentation and controls personnel working at GTRE33 informed the shift chemist that the incorrect radiation monitor was removed from service. The shift chemist then contacted the control room which had received Annunciator 61C, Process Rad Mon Fail. The shift chemist subsequently returned GTRE22 to service. With both trains of containment purge radiation monitors removed from service, TS 3.3.6, Condition A, was entered for having more than one train inoperable. The containment purge and supply dampers were immediately verified to be closed and remained closed with no containment purge in progress.

Analysis:

The inspectors determined that the failure to remove the correct containment radiation monitor from service in accordance with Procedure CHS AX-G02, Exchange of Radioactive Gaseous Monitor Particulate and Iodine Filters, is a performance deficiency.

Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRCs 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 is associated with the configuration control attribute for the Barrier Integrity Cornerstone; and it affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, this issue relates to the Containment Boundary preservation example of the configuration control attribute because the shift chemist removed both trains of safety-related radiation monitors from service for the same period of time without preplanning or ensuring a containment purge (i.e., a radiological release) was not in progress.

The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because both trains of the radiation monitor protective functions (i.e., to stop a containment purge on a high radiation signal) were affected but did not result in an actual open pathway in the containment barrier. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the shift chemist failed to apply appropriate human error prevention techniques such as self and peer-checks.

Enforcement:

Technical Specification 5.4.1.a requires procedures be established, implemented, and maintained covering, in part, the applicable activities specified in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Regulatory Guide 1.33 requires, in part, that maintenance procedures which can affect the performance of safety-related equipment should be properly performed in accordance with written procedures appropriate to the circumstances. Procedure CHS AX-G02, Exchange of Radioactive Gaseous Monitor Particulate and Iodine Filters, Revision 15, covers the stated activities for the safety-related containment purge exhaust Monitors GTRE22 and GTRE33. Procedure CHS AX-G02 Step 6.1.1.1 states: Upon verification that the applicable radiation monitor(s) are in bypass, proceed with Step 6.1.2. Contrary to the above, on February 20, 2007, both trains of containment purge exhaust radiation monitors, GTRE22 and GTRE33, were removed from service because the shift chemist did not correctly verify that the applicable radiation monitor was removed from service before proceeding with the maintenance. This issue and the corrective actions are being tracked by the licensee in Condition Report 2007-000661.

Because the violation was of very low safety significance and the issue was captured in the licensees corrective action program, this violation is being treated as a NCV consistent with Section VI.A of the NRC Enforcement Policy, NCV 05000482/2007002-07, failure to remove the correct containment radiation monitor from service.

4OA5 Other Activities

(Closed) Unresolved Item (URI)0500482/2006011-01, Potential Failure to Survey Discharges of Radioactive Material During the biennial radiation safety team inspection conducted August 14-18, 2006, the team identified a potential violation for failing to survey discharges of radioactive material.

During annual testing of the licensee's fire protection system, water obtained from the licensees cooling lake was discharged directly on to the ground without ensuring compliance with 10 CFR 20.2001(a). The licensees cooling lake contains radioactive material (tritium) at an average concentration of 13,000 picoCuries per liter. This radioactive material had previously been discharged and accounted for as an effluent from the plant.

This item was unresolved pending an NRC review of the applicable regulations to determine if the use of previously discharged radioactive material in effluents is considered licensed radioactive material that is subject to NRC regulations including disposal.

The NRC has determined that radioactive material properly released as an effluent is no longer considered licensed material, if subsequently used or possessed by a licensee, up to exempt concentration limits. Thus, this material is not subject to NRC regulations for the disposal of licensed radioactive material pursuant to 10 CFR 20.2001(a). Therefore, the licensees actions did not result in a violation of regulatory requirements.

This URI is closed.

4OA6 Meeting, Including Exit

Exit Meeting Summary

On March 2, 2007, the inspector presented the resolution of the URI 2006011-01 to Mr. S. E. Hedges, Senior Vice President, Operations, and other members of the licensee's staff who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

On March 29, 2007, the inspector presented the occupational radiation safety inspection results to Mr. S. Hedges, Vice President of Operations and Plant Manager, and other members of the licensee's staff who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

On April 11, 2007, the resident inspectors presented the inspection results to Mr. S. E. Hedges and other members of licensee management. Licensee management acknowledged the inspection findings. The inspectors identified that they had reviewed proprietary information but had returned it to licensee personnel.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. J. Garrett, Vice President, Engineering

S. E. Hedges, Vice President, Operations and Plant Manager

R. A. Muench, President and Chief Executive Officer

K. Scherich, Director, Engineering

M. Sunseri, Vice President, Oversight

P. Bedgood, Superintendent, Radiation Protection

T. Jensen, Superintendent, Chemistry

S. Koenig, Manager, Chemistry/Radiation Protection

B. Muilenburg, Licensing Engineer, Regulatory Affairs

M. Skiles, Supervisor, Radiation Protection

K. Thrall, Supervisor, Radiation Protection

P. Wagner, Steam Generator Engineer, Engineering

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000482/2007002-02 URI Use of wrench on manual service water valve handwheel (Section 1R13)

Opened and Closed

05000482/2007002-01 NCV Failure to implement fire protection impairment control permit requirements and compensatory measures (Section 1RO5)

05000482/2007002-03 NCV Failure to perform an adequate survey in a high radiation area (Section 2SO1)

05000482/2007002-04 NCV Failure to maintain sufficient records to furnish evidence of events significant to plant safety (Section 4OA3.1(1))

05000482/2007002-05 FIN Failure to maintain drawings technically accurate (Section 4OA3.1(2))

05000482/2007002-06 NCV Failure to maintain steady state reactor power at or below the licensed thermal power limit (Section 4OA3.2)

05000482/2007002-07 NCV Failure to remove the correct containment radiation monitor from service. (Section 4OA3.3)

Closed

05000482/2006011-01 URI Potential failure to survey discharges of radioactive material (Section 4OA5)