IR 05000305/2003006

IR 05000305-03-006; 07/01/03 - 09/30/03; Kewaunee Nuclear Power Plant; Equipment Alignment and Operability Determinations

ML033000156
Person / Time
Site:	Kewaunee
Issue date:	10/24/2003
From:	Louden P Division Reactor Projects III
To:	Coutu T Nuclear Management Co
References
IR-03-006
Download: ML033000156 (36)

Text

ber 24, 2003

SUBJECT:

KEWAUNEE NUCLEAR POWER PLANT NRC INTEGRATED INSPECTION REPORT 05000305/2003006

Dear Mr. Coutu:

On September 30, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Kewaunee Nuclear Power Plant. The enclosed integrated inspection report documents the inspection findings which were discussed on October 2, 2003, with Mr. Hoops and other members of your staff.

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

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

Based on the results of this inspection, there was one NRC-identified and one self-revealed finding of very low safety significance (Green). These findings were determined to involve violations of NRC requirements. However, because these violations were of very low safety significance, non-willful and non-repetitive, and because the violations were entered in your corrective program, the NRC is treating these issues as Non-Cited Violations, in accordance with Section VI.A.1 of the NRCs Enforcement Policy.

If you contest the subject or severity of a Non-Cited Violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region III, 801 Warrenville Road, Lisle, IL 60532-4351; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector Office at the Kewaunee facility. In accordance with 10 CFR 2.790 of the NRCs 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 (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Patrick L. Louden, Chief Branch 5 Division of Reactor Projects Docket No. 50-305 License No. DPR-43

Enclosure:

Inspection Report 05000305/2003006 w/Attachment: Supplemental Information

REGION III==

Docket No.: 50-305 License No.: DPR-43 Report No.: 05000305/2003006 Licensee: Nuclear Management Company, LLC Facility: Kewaunee Nuclear Power Plant Location: N 490 Highway 42 Kewaunee, WI 54216 Dates: July 1 through September 30, 2003 Inspectors: R. Krsek, Senior Resident Inspector J. Cameron, Project Engineer M. Maley, Reactor Engineer, NRR B. Jorgensen, Reactor Engineer D. McNeil, Reactor Engineer D. Chyu, Reactor Engineer Observers: R. Berg, Resident Inspector Approved By: Patrick Louden, Chief Branch 5 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000305/2003006; 07/01/2003 - 09/30/2003; Kewaunee Nuclear Power Plant; Equipment

Alignment and Operability Determinations.

This report covers a 3-month period of baseline resident and announced licensed operator requalification program inspections. The inspections were conducted by the resident inspectors, and Region III and NRC Headquarters inspectors. The inspection identified two Green findings which were Non-Cited Violations (NCVs). One NRC-identified Green finding and one self-revealed Green finding associated with NCVs were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, Significance Determination Process (SDP). 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.

A. Inspector-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to prescribe instructions or procedures appropriate to the circumstances for the seismic control of equipment stored near the vicinity of the A Auxiliary Feedwater (AFW) piping to the A Steam Generator, an activity affecting quality. The inspectors identified during plant walkdowns that following the 2003 Refueling Outage, portable plant equipment, including two portable 2.5-ton cranes, were stored in close proximity to the AFW piping, without the use of seismic restraints.

This inspector-identified finding was greater than minor because if left uncorrected the finding would become a more significant safety concern. The failure to ensure that equipment stored near the safety-related A AFW piping was seismically restrained affected the mitigating systems attributes of configuration control and protection against external factors (seismic). In addition, the finding affected the cornerstone objective of ensuring the reliability and capability of the engineered safeguards systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance mainly because the finding did not represent an actual loss of safety function of a system and did not screen as potentially risk significant due to a seismic event. Therefore the finding screened as Green utilizing the Inspection Manual Chapter 0609 Significance Determination Process Phase 1 Worksheet for Mitigating Systems. (Section 1R04.1)

Green.

A Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was self-revealed when the licensee, in preparing and verifying the response to NRC

Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors, dated June 9, 2003, determined that the containment refueling cavity standpipe had not been installed after the Spring 2003 Refueling Outage. A procedure revision, issued prior to the 2003 Outage, had removed prescribed instructions to install the refueling cavity drain standpipe following reactor vessel refueling activities. The inspectors also concluded that this finding had, as a primary cause, a human performance deficiency.

This self-revealed finding was greater than minor because the failure to ensure that the refueling cavity standpipe was installed affected the mitigating systems attributes of configuration control and procedure quality. In addition, the finding affected the cornerstone objective of ensuring the reliability and capability of the engineered safeguards systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance mainly because the finding did not represent an actual loss of safety function of a system. Therefore the finding screened as Green utilizing the Inspection Manual Chapter 0609 Significance Determination Process Phase 1 Worksheet for Mitigating Systems. (Section 1R15.1)

B. Licensee-Identified Violation No findings of significance were identified.

REPORT DETAILS

Summary of Plant Status

The plant was operated at or near full power for most of the inspection period.

Exceptions included a planned power reduction to approximately 70 percent power for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> on July 3 and 4, 2003, to perform routine turbine valve and auxiliary feedwater pump testing.

On July 16, 2003, the licensee received a license amendment to increase reactor power to 1673 Megawatts-Thermal, through a Measurement Uncertainty Recapture Uprate. Therefore, from July 16 through July 27, 2003, full reactor power was reported as 99 percent since the licensee had not yet implemented the power uprate. Finally, on July 24 through 26, 2003, plant power was reduced to 98.5 percent to perform Reactor Protection System calibration and testing to support implementation of the Measurement Uncertainty Recapture Uprate.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R04 Equipment Alignment

.1 Auxiliary Feedwater (AFW) System Train A Piping

a. Inspection Scope

The inspectors conducted a partial walkdown of the A Train of Auxiliary Feedwater while the B Train of AFW equipment was out-of-service which also included a verification walkdown of the AFW Train B once it was returned to service. One inspection procedure sample was completed. The inspectors verified that the systems were correctly aligned to perform their design safety function.

In preparation for the walkdowns, the inspectors reviewed the system lineup checklists, normal operating procedures, abnormal and emergency operating procedures, and system drawings to verify the correct system lineup. During the walkdowns, the inspectors also examined valve positions and electrical power availability to verify that valve and electrical breaker positions were consistent with, and in accordance with, the licensees procedures and design documentation. The material condition of the equipment was also inspected. Finally, the inspectors verified seismic housekeeping requirements were in place for areas surrounding the AFW piping and equipment.

b. Findings

Introduction:

The inspectors identified a Green finding associated with a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to prescribe instructions or procedures appropriate to the circumstances for the seismic control of equipment stored near the vicinity of the A AFW piping to the A Steam Generator, an activity affecting quality. In addition, the inspectors identified during plant walkdowns that following the 2003 Refueling Outage, portable plant equipment, including two portable 2.5-ton cranes, were stored in close proximity to the auxiliary feedwater piping without the use of seismic restraints.

Description:

On July 24, 2003, the inspectors identified during an equipment walkdown of the A Train of the AFW system, that heavy mobile equipment was stored adjacent to the AFW system piping and was not secured. At the time, two 2.5-ton portable cranes were stored within two inches of the 3-inch A Train AFW piping. The inspectors noted that the AFW piping penetrated the floor of the Mechanical Equipment Room on the 606-foot elevation of the Auxiliary Building and then traversed horizontally several feet to the containment penetration for the A AFW piping to the A Steam Generator. The inspectors also noted that a 3/4-inch branch line utilized for chemical addition was connected to the AFW piping in the same area.

The inspectors reviewed licensee procedure GNP-01-31.01, Plant Cleanliness and Storage, Revision E, dated October 17, 2002, and determined that the procedure prescribed the area surrounding the A Train AFW piping as an acceptable storage area, not subject to seismic controls for the storage of equipment near safety-related equipment. In further discussions with the licensee, the inspectors determined that the procedure was based on a September 10, 1996, licensee analysis which utilized Electric Power Research Institute Document NP-6041-SL, A Methodology for Assessment of Nuclear Plant Seismic Margin (Revision 1). After consultation with NRC Region III Inspectors and Nuclear Reactor Regulation Technical Staff, the inspectors concluded that the Electric Power Research Institute Document had not been endorsed by the NRC for plant licensing basis evaluations, and since a seismic event is a design basis event, the use of this document may not be appropriate to the circumstances.

The licensee concurred with the inspectors observations and initiated Condition Report CAP017393, Potential Interaction of three-inch AFW Piping with Portable Equipment.

In addition, the licensee took immediate corrective actions to initially secure and later remove the portable equipment stored in close proximity to the safety related A Train AFW piping. The licensee also established a temporary exclusion zone surrounding the piping to ensure no future equipment was stored in the area. The licensees pending corrective actions included the determination of a safe distance from the piping for storage of equipment in the Mechanical Equipment Room and implementation of permanent changes to Procedure GNP-01-31.01.

Analysis:

The inspectors determined that the failure to have appropriate instructions or prescribed actions in Procedure GNP 01.31.01 to ensure that the equipment stored near the A AFW piping would not adversely affect the function of the AFW piping during a design basis seismic event was considered a licensee performance deficiency warranting a significance evaluation. This inspector-identified finding was greater than minor because if left uncorrected the finding would become a more significant safety concern. The failure to ensure that equipment stored near the safety-related A AFW piping was seismically restrained affected the mitigating systems attributes of configuration control and protection against external factors (seismic). In addition, the finding affected the cornerstone objective of ensuring the reliability and capability of the engineered safeguards systems that respond to initiating events to prevent undesirable consequences.

The inspectors evaluated the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding:

• was not a design or qualification deficiency;

• did not represent an actual loss of safety function of a system;

• did not represent an actual loss of a safety function of a single train for greater than Technical Specification outage time;

• did not represent an actual loss of a safety function of one or more Non-Technical Specification trains of equipment designated as risk significant;

• did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event; C did not involve the total loss of any safety function that contributed to core damage accident sequences initiated by seismic events; and C did not involve the loss or degradation of equipment or function designed to mitigate a seismic initiating event.

Therefore, the finding was determined to be of very low safety significance (Green).

Enforcement:

10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances.

Contrary to this requirement, Procedure GNP 01.31.01, Plant Cleanliness and Storage, Revision E, did not provide instructions or procedures to ensure that the equipment stored near the A AFW piping would not adversely affect the function of the AFW piping during a design basis seismic event, an activity affecting quality; therefore, the procedure was not appropriate to the circumstances. The inspectors determined this finding was a violation of 10 CFR Part 50, Appendix B, Criterion V. Because this violation was of very low safety significance (Green) and was documented in the licensees corrective action program as CAP017393, this finding is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy.

(NCV 05000305/2003006-01)

.2 Other Partial Equipment Alignments

a. Inspection Scope

The inspectors conducted partial walkdowns of the system trains listed below while the opposite train of equipment was out-of-service or after return-to-service of a system train to verify that the systems were correctly aligned to perform their design safety function. In preparation for the walkdowns, the inspectors reviewed the system lineup checklists, normal operating procedures, abnormal and emergency operating procedures, and system drawings to verify the correct system lineup. During the walkdowns, the inspectors also examined valve positions and electrical power availability to verify that valve and electrical breaker positions were consistent with, and in accordance with, the licensees procedures and design documentation. The material condition of the equipment was also inspected.

Partial system walkdowns were conducted on the following systems which constituted three inspection procedure samples:

• Diesel Generator 1A;

• Safety Injection Pump 1A; and

• Service Water Pumps Trains A and B.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of fire fighting equipment, the control of transient combustibles and ignition sources, and on the condition and operating status of installed fire barriers. The inspectors selected fire areas for inspection based on their overall contribution to internal fire risk, as documented in the Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate a plant transient, or their impact on the plants ability to respond to a security event. The inspectors used the documents listed in Attachment 1 to verify 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 that fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program.

The following 7 areas were inspected by walkdowns:

• Fire Zone TU-91, Emergency Diesel Generator (EDG)1A Day Tank Room;

• Fire Zone TU-93, EDG 1B Day Tank Room;

• Fire Zone TU-95C, AFW Pump Room 1A;

• Fire Zone AX-23D, Component Cooling Water Pump 1B Room;

• Fire Zone SC-70B, Screenhouse, South Area;

• Fire Zone TC-101, Technical Support Center (TSC) 606' Level; and

• Fire zone TC-102, TSC Non-Safeguards Battery and Electric Equipment Room.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

a. Inspection Scope

On August 7, 2003, the licensee conducted heat exchanger performance monitoring for the 1B EDG service water support systems. The inspectors reviewed the test procedure and reviewed the test data to verify: 1) that the test was performed as written; 2) that the acceptance criteria were adequate to demonstrate acceptable heat transfer capability of the heat exchanger; and 3) that the test data met the acceptance criteria.

The inspectors also verified that the test accounted for instrument inaccuracies and that the test frequency was sufficient to provide early detection of heat exchanger degradation prior to any loss of heat removal capabilities below design values. Finally, the inspectors compared the current test results with previous test data to verify the performance of the heat exchangers tested. This activity constituted one inspection procedure sample.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

.1 Written Examination and Operating Test Results

a. Inspection Scope

The inspectors reviewed the pass/fail results of individual written tests, operating tests, and simulator operating tests (required to be given per 10 CFR 55.59(a)(2))

administered by the licensee during calender year 2003. This represents one sample.

b. Findings

No findings of significance were identified.

.2 Resident Inspector Quarterly Inspection of Licensed Operator Requalification

a. Inspection Scope

On September 30, 2003, the inspectors observed a simulator dynamic requalification examination for Cycle 03-05 to evaluate crew performance, clarity and formality of communications, ability to take timely actions in a safe direction, procedure use, control board manipulations, oversight and direction from supervisors, group dynamics and annunciator response. Additionally, the inspectors evaluated the crews implementation of the facilitys abnormal and emergency operating procedures, oversight and direction provided to the crew by the shift manager and control room supervisor.

The inspectors also compared the simulator board configuration with the actual control room board configuration to verify that the simulator environment matched the actual control room environment as closely as possible. The inspectors observed the post-scenario critiques to determine whether performance issues were accurately identified and addressed by the licensee. This observation constituted one quarterly inspection procedure sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the implementation of the maintenance rule for the systems and/or equipment problems listed below to verify that component and equipment failures were identified, entered, and scoped within the Maintenance Rule. The inspectors also verified that the system or equipment was properly categorized and classified as (a)(1)or (a)(2) in accordance with 10 CFR 50.65. The inspectors reviewed a sample of station logs, maintenance work orders, action requests, functional failure evaluations, unavailability records, and corrective action reports to verify that the licensee was identifying issues related to the Maintenance Rule at an appropriate threshold and that corrective actions were appropriate. Additionally, the inspectors reviewed the licensees performance criteria to verify that the criteria adequately monitored equipment performance. The systems reviewed by the inspectors, which constituted two quarterly inspection procedure samples, were the:

• Emergency Diesel Generator Systems (excluding the TSC Diesel Generator)

• Safety Injection Systems.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessment and Emergent Work Evaluation

a. Inspection Scope

The inspectors reviewed the licensees evaluation and assessment of plant risk, scheduling, and configuration control during the planned and emergent work activities listed below. In particular, the licensees planning and management of maintenance was evaluated to verify that on-line risk was acceptable and in accordance with the requirements of 10 CFR 50.65(a)(4).

Additionally, the inspectors compared the assessed risk configuration against the actual plant conditions and any in-progress evolutions or external events to verify that the assessment was accurate, complete, and appropriate. Licensee actions to address increased on-line risk during these periods were also inspected to verify that actions were in accordance with approved administrative procedures. The inspectors conducted the following five inspection procedure samples during the inspection period:

• Safety Monitor Risk Assessment for July 7 through 11, 2003;

• Safety Monitor Risk Assessment for July 14 through 18, 2003;

• Safety Monitor Risk Assessment for July 21 through 25, 2003;

• Safety Monitor Risk Assessment for August 18 through 22, 2003; and

• Safety Monitor Risk Assessment for September 22 through 26, 2003.

b. Findings

No findings of significance were identified.

1R14 Personnel Performance During Non-Routine Plant Evolutions

a. Inspection Scope

The inspectors observed licensee personnel performance during the planned plant evolutions associated with the implementation of the Measurement Uncertainty Recapture Uprate which occurred the last 2 weeks of July 2003. The inspectors observed operators performance during the scheduled testing and calibration of Reactor Protection System Channels, which were new evolutions.

The inspectors reviewed operator log entries, and plant process computer data to verify the appropriate plant response during the implementation of power uprate and to ensure plant performance was consistent with the expected changes in operating parameters.

The inspectors also observed briefings conducted for the plant operators related to the implementation of power uprate, and verified operations procedures and training were updated to reflect the changes associated with the power uprate.

Finally, the inspectors verified that the licensee identified problems associated with the power uprate implementation at the appropriate thresholds and that conditions adverse to quality were entered into the licensees corrective action system. This activity constituted one inspection procedure sample.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

.1 Reactor Cavity Drain Standpipe

a. Inspection Scope

The inspectors reviewed design basis and Updated Final Safety Analysis Report information, and technical specification requirements to verify the technical adequacy of the operability evaluation performed for the failure to install the Reactor Cavity Drain Pipe. The inspectors also verified that the Engineered Safeguards System operability was properly justified and that the system remained available, such that no unrecognized increase in risk occurred. This activity constituted one inspection procedure sample.

b. Findings

Introduction:

A Green finding associated with a NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was self-revealed when the licensee, in preparing and verifying the response to NRC Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors, dated June 9, 2003, determined that the containment refueling cavity standpipe had not been installed after the Spring 2003 Refueling Outage. A procedure revision, issued prior to the 2003 Refueling Outage, had removed prescribed instructions to install the refueling cavity drain standpipe following reactor vessel refueling activities.

Description:

On June 9, 2003, the NRC issued Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors, to all holders of operating licenses for pressurized-water reactors. The Bulletin requested licensees to review the potential impact of debris blockage on the Emergency Core Cooling System and Containment Spray System recirculation functions. While reviewing and validating the response to the NRC Bulletin, the licensee determined that the containment refueling cavity drain standpipe may not have been installed following the Spring 2003 Refueling Outage.

The function of the refueling cavity drain standpipe was to prevent and minimize the transport of debris, mainly unqualified containment coatings, from entering Containment Sump A following a design basis Large Break Loss of Coolant Accident (LBLOCA).

The transport path of concern was that debris which enters Containment Sump A could be transported to the containment basement floor and Containment Sump B when Containment Sump A overflows, as designed, following a LBLOCA. Debris transported to Containment Sump B was a sump screen blockage concern, in that, the Residual Heat Removal, Safety Injection and Internal Containment Spray systems take suction from Containment Sump B during the Recirculation Phase of a LBLOCA.

During a routine, at-power containment entry on August 13, 2003, the licensee verified that the refueling cavity drain standpipe was not installed and was positioned in the lower reactor cavity in containment. The licensee performed an operability determination and determined that containment sump recirculation system was considered operable, but non-conforming in accordance with NRC Generic Letter 91-18, based on the following:

• Engineering Calculation C10922, concluded that eight cubic feet of unqualified coatings in containment would not result in inoperability of the emergency core cooling systems. Currently, there existed approximately three cubic feet of unqualified coatings in containment, of which only 1.4 cubic feet would potentially enter the reactor cavity post accident; and

• The transport path for the unqualified containment coatings from the reactor cavity to Containment Sump B was torturous, in that a number of debris diversions were present.

The inspectors noted a minor issue, in that, the initial licensee operability determination did not address the potential fluid blockage which could be caused by the refueling cavity standpipe, if positioned near the refueling cavity drain. The licensee subsequently determined that the shape and orientation of the standpipe was such that, even if positioned near the refueling cavity drain, fluid transport from the refueling cavity to Containment Sump A would not be negatively impacted.

The licensee determined that the procedure steps prescribing the installation of the refueling cavity drain standpipe were inadvertently deleted in a 1998 revision of Procedure RF-1.0, KNPP Refueling Procedure, due to the belief that the refueling cavity drain standpipe was permanently installed plant equipment. The licensees immediate corrective actions included adding the installation of the refueling cavity drain standpipe to the licensees forced outage work scope, in addition to initiating a procedure change to correct Procedure RF-1.0.

Analysis:

The inspectors determined that the licensees failure to have appropriate instructions or prescribed actions in Procedure RF-01.00, to ensure that the containment refueling cavity drain standpipe was installed following refueling activities was considered a licensee performance deficiency warranting a significance evaluation. This self-revealed finding was greater than minor because the failure to ensure that the refueling cavity standpipe was installed affected the mitigating systems attributes of configuration control and procedure quality. In addition, the finding affected the cornerstone objective of ensuring the reliability and capability of the engineered safeguards systems that respond to initiating events to prevent undesirable consequences.

The inspectors evaluated the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding:

• was not a design or qualification deficiency;

• did not represent an actual loss of safety function of a system;

• did not represent an actual loss of a safety function of a single train for greater than Technical Specification outage time;

• did not represent an actual loss of a safety function of one or more Non-Technical Specification trains of equipment designated as risk significant; and

• did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.

Therefore, the finding was determined to be of very low safety significance (Green).

The inspectors also determined that the finding affected the cross-cutting area of Human Performance, because the procedure error contained in Refueling Procedure RF-01.00 was caused by the removal of a procedure step without the appropriate verification and validation of the actual plant configuration.

Enforcement:

10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances.

Contrary to this requirement, Procedure RF-01.00, KNPP Refueling Procedure, Revision J, did not provide instructions or procedures to ensure that the containment refueling cavity drain standpipe was reinstalled following reactor refueling activities; therefore, the procedure was not appropriate to the circumstances. The inspectors determined this finding was a violation of 10 CFR Part 50, Appendix B, Criterion V.

Because this violation was of very low safety significance (Green) and was documented in the licensees corrective action program as CAP017357, this finding is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy.

(NCV 05000305/2003006-02)

.2 Other Operability Determinations Reviewed

a. Inspection Scope

The inspectors reviewed design basis information and technical specification requirements to verify the technical adequacy of the operability evaluations listed below and to verify that system operability was properly justified and that the system remained available, such that no unrecognized increase in risk occurred.

The inspectors reviewed the following corrective action process (CAP) operability evaluations, which constituted four inspection procedure samples:

• CAP017169, A Emergency Diesel Generator Speed Cycling;

• CAP017273, Blue Channel Reactor Protection System T-Average Response;

• CAP017771, Generic Letter 89-13 Test Methodology Questionable for the Emergency Diesel Generator Coolers and Containment Fan Coil Units; and

• CAP017736, Quality Assurance Typing on Fire Protection Relays for the A and B Emergency Diesel Generator Start and Trip circuit Not Correct.

b. Findings

No findings of significance were identified.

1R16 Operator Workarounds

a. Inspection Scope

The inspectors reviewed previously identified operator workarounds, equipment deficiency logs, and control room deficiencies to verify that the cumulative effects did not create significant adverse consequences regarding the reliability, availability and operation of accident mitigating systems. The inspectors also assessed these cumulative effects on the ability to implement abnormal and emergency response procedures in a correct and timely manner.

The inspectors reviewed the planned actions to address operator workarounds to verify that the priority to resolve the deficiencies were appropriate when considering the potential impact on plant risk and safety. In addition, the inspectors reviewed emergent risk significant operator workarounds to determine if the functional capability of a system or human reliability of an initiating event was affected. Finally, the inspectors reviewed condition reports regarding operator workarounds to verify that the corrective actions were prioritized and appropriate, commensurate with the safety significance of the issue.

These activities constituted one inspection procedure sample.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the post-maintenance testing activities associated with the scheduled and emergent work activities listed below to verify that the testing was adequate for the scope of the maintenance work performed. The inspectors reviewed the tests acceptance criteria to ensure that the criteria was clear and that testing demonstrated operational readiness consistent with the design and licensing basis documents.

The inspectors attended pre-job briefings, when possible, to verify that the impact of the testing had been properly characterized; and observed or reviewed the test to verify that the test was performed as written and that all testing prerequisites were satisfied.

Following the completion of each test, the inspectors completed walkdowns of the affected equipment, when applicable, to verify that the test equipment was removed and that the equipment was returned to a condition in which it could perform its safety function. The inspectors also reviewed the completed test data to ensure the test acceptance criteria were met for the following activities, which constituted seven inspection procedure samples:

• Replacement of 1B Safety Injection Pump Seal Cooling Heat Exchanger;

• Partial Installation of A Component Cooling Water Pump Recirculation Piping;

• Partial Installation of B Component Cooling Water Pump Recirculation Piping;

• Direct Current Supply & Distribution 7.5 KVA Inverter BRA-111 Electrical Maintenance;

• Component Cooling Water System Minimum Flow Recirculation Line Testing and Data Gathering Activities;

• Service Water System - Motor-Operated Valves SW1300 and SW1306, Service Water Outlet and Bypass Valves for B Component Cooling Water Heat Exchanger, preventive maintenance; and

• Service Water System - Motor Operated Valve SW601B, Emergency Service Water Suction for AFW Pump 1B, preventive maintenance.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed and reviewed the surveillance testing results for the surveillances listed below to verify that the equipment was capable of performing the intended safety function and that the surveillance tests satisfied the requirements contained in Technical Specifications and the licensees procedures. The inspectors reviewed the surveillance tests to verify the tests were adequate to demonstrate operational readiness consistent with the design and licensing basis documents, and that the testing acceptance criteria were well documented and appropriate to the circumstances.

Portions of the test were observed to verify the test was performed as written, that all testing prerequisites were satisfied, and that the test data was complete, appropriately verified, and met the requirements of the testing procedure. Following the completion of the tests, when applicable, the inspectors conducted walkdowns of the affected equipment to verify that the test equipment was removed and that the effected equipment was returned to an operable condition.

The inspectors observed and reviewed the following tests, which constituted seven inspection procedure samples:

• Surveillance Procedure SP-42-312B, Diesel Generator B Start-Up Air Leakage and Monthly Availability Testing;

• Surveillance Procedure SP-23-100A, Train A Internal Containment Spray Pump and Valve Inservice Testing;

• Surveillance Procedure SP-31-168A, Component Cooling Water Pump Train A Inservice Testing;

• Repetitive Test Procedure RT-DGM-10, Technical Support Center Diesel Generator Monthly Test;

• Surveillance Procedure SP-23-100B, Train B Internal Containment Spray Pump and Valve Inservice Testing;

• Surveillance Procedure SP-34-099A, Train A Residual Heat Removal Pump and Valve Inservice Testing; and

• Surveillance Procedure SP-18-043, Containment Pressure Instrument Monthly Channel Testing.

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the modification documentation and associated 10 CFR 50.59 evaluation for the following temporary plant modification:

• TCR-03-26, Power Operated Relief Valve Discharge Temperature Element The inspectors verified that the temporary modification did not adversely impact other safety-related equipment and that the modification was being controlled in accordance with the licensees administrative procedures. The inspectors also verified that the modification did not affect system operability or availability.

In addition, the inspectors reviewed condition reports to verify that temporary modifications problems were entered into the corrective action program with the appropriate significance characterization. This activity constituted one inspection procedure sample.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

The inspectors observed the emergency response activities associated with the licensee emergency preparedness drill conducted on August 6, 2003. The drill was designed to exercise the licensees onsite and offsite emergency response organizations and emergency plans. The inspectors observed portions of the drill from the control room simulator, the TSC and the Operations Support Facility to evaluate the licensees evaluation, classification, and notification of the simulated event. The inspectors also attended both the drill controllers debrief and the general drill critique to determine whether the licensee properly identified drill performance weaknesses at an appropriate threshold.

The inspectors verified that the weaknesses observed during the emergency preparedness drill were identified during the licensees critique and captured in the corrective action program. This activity constituted one inspection procedure sample.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

.1 Reactor Safety Strategic Area

a. Inspection Scope

The inspectors reviewed the licensees Performance Indicator data collection process and historical data from the third quarter of 2002 through the second quarter of 2003 to verify the accuracy of licensee collected and submitted data. The inspectors used performance indicator definitions and guidance contained in Revision 2 of the Nuclear Energy Institutes Document 99-02, Regulatory Assessment Performance Indicator Guideline, to verify the performance indicator data. The following Performance Indicators were evaluated which constituted four inspection procedure samples:

• Unplanned Power Changes;

• Emergency Alternating Current Power System Unavailability;

• Auxiliary Feedwater Heat Removal System Unavailability; and

• Residual Heat Removal System Unavailability.

The inspectors reviewed corrective action documents, monthly operating reports, completed surveillance procedures, control room logs, and licensee event reports to independently verify the data that the licensee had collected. The inspectors also independently re-performed calculations for system unavailability when applicable.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that issues were entered into the licensees corrective action system at an appropriate threshold, that adequate attention was given to timely corrective actions, and that adverse trends were identified and addressed. The inspectors also reviewed all condition reports written by licensee personnel during the inspection quarter. Minor issues entered into the licensees corrective action system as a result of inspectors observations are included in the list of documents reviewed which is attached to this report.

b. Findings

No findings of significance were identified.

4OA3 Event Followup

.1 May 1, 2003 Event Notification Number 39816: Loss of Radiation Level Indications in

the Auxiliary Building and Containment Vent Stacks.

On May 1, 2003, the licensee conservatively reported a major loss of emergency assessment capability under 10 CFR 50.72(b)(3)(xiii). The licensees report stated the following, in part: it was determined that the Containment and Auxiliary Building SPING Units were not functioning properly and could not be relied upon to determine Emergency Classifications based on Emergency Action Level Chart A(2). Alternate Radiation Monitors remain available to provide indication of radiation levels in the Auxiliary Building and Containment Vent Stacks; however, their use is not currently proceduralized. The use of alternate radiation monitors to determine the Emergency Classification could impair the plants ability to declare an emergency in a timely manner.

During routine followup of this event notification, the inspectors determined that the licensee had written a condition report to document the equipment deficiencies and taken immediate contingency actions at the time the event notification was made.

However, the licensee had not assessed the impact on the emergency plan with these radiation monitors being out-of-service from approximately November 2002 through April 2003 and with no compensatory actions in place. Licensee management concurred that this issue needed further evaluation and initiated Condition Report CAP018205 to evaluate the consequences of the radiation monitors being out-of-service from approximately November 2002 through April 2003. Therefore, pending the licensees evaluation of the circumstances surrounding this issue, the inspectors have identified this issue as an Unresolved Item, URI 05000305/2003006-03.

.2 August 14, 2003, Loss of the East Coast Electric Power Grid: On August 14, 2003,

shortly after 3:00pm (Central Time) there was a loss of the east coast power grid.

Shortly after the grid loss, the inspectors observed onsite plant parameters and status, and reviewed the control room alarms and conditions immediately prior to and following the east coast event. The inspectors verified that plant equipment responded as expected and that the licensee entered minor equipment issues observed shortly after this evolution into the corrective action program. The inspectors verified that no significant effects were seen at the plant as a result of this event.

.3 (Closed) Licensee Event Report (LER) 50-305/2001-002-03: Non-Rated Fire Barrier

Separating Redundant Appendix R Safe Shutdown Capabilities.

The licensee submitted an update to the original Licensee Event Report to reflect the licensees completed Root Cause Evaluation and additional corrective actions taken in response to this event. This issue was initially discussed as an Unresolved Item URI 50-305/01-02-1 and subsequently closed as an Non-Cited Violation in NRC Inspection Report 50-305/01-11. A final safety significance determination of was documented in a letter to the licensee dated June 6, 2001. The inspectors concluded that this update provided supplemental information regarding this issue based on the licensees completion of a Root Cause Evaluation and the updated information did not affect the final safety significance documented. Therefore, the inspectors considered this item closed.

4OA4 Cross-Cutting Aspects of Findings

.1 A Green finding described in Section 1R15.1 of this report had, as a primary cause, a

human performance deficiency, in that, a revision to Refueling Procedure RF-01.00 removed the procedure step to ensure the refueling cavity drain standpipe was installed following refueling activities without the appropriate verification and validation of the actual plant configuration.

4OA5 Other Activities

(Closed) Unresolved Item (URI) 50-305/01-02-02: Relay room carbon dioxide system testing. This issue was reviewed by the Office of Nuclear Reactor Regulation and the Office of General Counsel. The NRC staffs acceptance of the Kewaunee system was based on the licensees commitment to satisfy the requirements of National Fire Protection Association (NFPA) 12-1973, Carbon Dioxide Extinguishing Systems.

Based on the licensees commitment to meet the NFPA 12-1973, the licensing basis requirement for carbon dioxide concentration was 50 percent and because NFPA 12-1973 did not require a specific soak time, there was no specific licensing basis soak time for Kewaunee. Since the 1973 edition of the standard did not explicitly require a full discharge test, no full discharge test was required. The NRC concluded that the licensing basis for the gaseous suppression system at Kewaunee has been met. This item is closed.

4OA6 Meetings

On October 2, 2003, the resident inspectors presented the inspection results to Mr. Hoops and other members of his staff, who acknowledged the findings. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

.2 Interim Exit Meetings

An exit meeting was conducted for:

• Results of Licensed Operator Requalification Testing for Calender Year 2003 and Applicability of NRC Inspection Manual Chapter 0609, Appendix I, Operator Requalification Human Performance Significance Determination Process (SDP),with K. Davison on August 26, 2003.

4OA7 Licensee-Identified Violation

None.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Nuclear Management Company, LLC

T. Coutu, Site Vice President

K. Hoops, Site Director

G. Arent, Regulatory Affairs

L. Gerner, Acting Regulatory Affairs Manager

L. Armstrong, Engineering Director

K. Davison, Operations Manager

S. Baker, Manager, Radiation Protection

M. Fencl, Security Manager, Kewaunee/Point Beach

G. Harrington, Licensing

J. McCarthy, Plant Manager

B. Presl, NMC Security Consultant

S. Putman, Assistant Plant Manager, Maintenance

R. Repshas, Manager, Site Services

J. Riste, Licensing Supervisor

J. Stafford, Superintendent, Operations

NRC Personnel

J. Lamb, Project Manager

T. McMurtray, Acting Project Manager

J. Gavula, Senior Reactor Engineer

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000305/2003006-01 NCV Green 10 CFR 50, Appendix B, Criterion V NCV for the Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Seismic Storage of Equipment Near the A Auxiliary Feedwater Piping (Section 1R04.1)

05000305/2003006-02 NCV Green 10 CFR 50, Appendix B, Criterion V NCV for the Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Installation of the Refueling Cavity Drain Standpipe Following Refueling Activities (Section 1R15.1)

05000305/2003006-03 URI Licensee Evaluation of Circumstances Surrounding Event Notification 39816, dated May 1, 2003 (Section 4OA3.1)

Closed

05000305/2003006-01 NCV Green 10 CFR 50, Appendix B, Criterion V NCV for the Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Seismic Storage of Equipment Near the A Auxiliary Feedwater Piping (Section 1R04.1) (Section 1R04.1)

05000305/2003006-02 NCV Green 10 CFR 50, Appendix B, Criterion V NCV for the Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Installation of the Refueling Cavity Drain Standpipe Following Refueling Activities (Section 1R15.1) (Section 1R15.1)

05000305/2001002-02 URI Relay room carbon dioxide system testing.

(Section 4OA5)

05000305/2001002-03 LER Non-Rated Fire Barrier Separating Redundant Appendix R Safe Shutdown Capabilities (Section 4OA3.3)

Discussed

None Attachment