IR 05000255/2003006

IR 05000255-03-006; on 07/01/2003 - 09/30/2003; Palisades Nuclear Generating Plant; Operability Evaluations; Radiation Protection

ML033030565
Person / Time
Site:	Palisades
Issue date:	10/30/2003
From:	Eric Duncan NRC/RGN-III/DRP/RPB6
To:	Domonique Malone Nuclear Management Co
References
IR-03-006
Download: ML033030565 (51)

Text

ber 30, 2003

SUBJECT:

PALISADES NUCLEAR GENERATING PLANT NRC INTEGRATED INSPECTION REPORT 05000255/2003006

Dear Mr. Malone:

On September 30, 2003, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Palisades Nuclear Generating Plant. The enclosed report documents the inspection findings which were discussed on October 9, 2003, with you 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, four findings of very low safety significance (Green)

were identified, which were determined to involve violations of NRC requirements. However, because these violations were of very low safety significance and because they have been entered into your corrective action program, the NRC is treating these violations 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 a copy 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 Resident Inspector Office at the Palisades facility. In accordance with 10 CFR 2.790 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 (PARS) component of NRC's 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/

Eric R. Duncan, Chief Branch 6 Division of Reactor Projects Docket No. 50-255 License No. DPR-20

Enclosure:

Inspection Report 05000255/2003006 w/Attachment: Supplemental Information

REGION III==

Docket No: 50-255 License No: DPR-20 Report No: 05000255/2003006 Licensee: Nuclear Management Company, LLC Facility: Palisades Nuclear Generating Plant Location: 27780 Blue Star Memorial Highway Covert, MI 49043-9530 Dates: July 1 through September 30, 2003 Inspectors: J. Lennartz, Senior Resident Inspector M. Garza, Resident Inspector R. Alexander, Radiation Specialist Inspector Approved by: E. R. Duncan, Chief Branch 6 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000255/2003006; 07/01/2003 - 09/30/2003; Palisades Nuclear Generating Plant;

Operability Evaluations; Radiation Protection.

This report covers a 3-month period of baseline resident inspections and a routine baseline radiation protection inspection. The inspections were conducted by the resident inspectors and a regional radiation specialist inspector. Four Green findings with associated Non-Cited Violations (NCVs) were identified during the inspection. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC)0609, Significance Determination Process, (SDP). Findings for which the SDP 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: Barrier Integrity

Green.

A finding of very low safety significance was self-revealed when the Containment Air Cooler Fan V-4A motor bearing failed and the fan tripped unexpectedly on July 1, 2003, after the fan was declared operable and returned to service following emergent repairs on June 20, 2003. A lack of rigor in the technical evaluation to determine the operability for Fan V-4A on June 20 resulted in the fan being declared operable and returned to service with more significant motor bearing degradation than recognized by licensee personnel. The primary cause of this finding was related to the cross-cutting area of Problem Identification and Resolution.

The finding was more than minor because the finding was associated with the Human Performance attribute of the barrier integrity cornerstone and adversely impacted the cornerstone objective to provide reasonable assurance that the containment barrier protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because there was no adverse impact on the physical integrity of reactor containment and there was no adverse impact on the atmospheric pressure control function of the reactor containment. Corrective actions to address the issue included replacing the motor for Fan V-4A and entering all containment air cooler fans and motors into a predictive maintenance program. One Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified.

(Section 1R15)

Cornerstone: Occupational Radiation Safety

Green.

A finding of very low safety significance was self-revealed when two workers entered a high radiation area to move a drum and trash bags of radioactive material out of the area without obtaining a briefing regarding the radiological conditions in the area.

The issue was associated with the Human Performance attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material because the workers were not sufficiently cognizant of the radiation fields they could have encountered while inside the high radiation area. The finding was of very low safety significance because the radiological conditions the workers could have encountered were not sufficient to produce a substantial potential for an exposure in excess of regulatory limits. To address this issue, the individuals involved were administratively precluded from entering the Radiologically Controlled Area for the remainder of the outage. Additionally, training to reinforce radiation protection standards and expectations was provided to radiation workers. One Non-Cited Violation for the failure to meet the requirements of Technical Specification 5.7.1.e for the conduct of pre-entry high radiation area briefings was identified. (Section 2OS1.5)

Green.

A finding of very low safety significance was self-revealed when a worker failed to stop work and contact radiation protection personnel upon receiving an electronic dosimetry dose rate alarm while rigging a drum of radioactive material to be removed from a posted high radiation area.

The issue was associated with the Human Performance attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material because the failure to appropriately act upon hearing the alarm was a failure of the radiation safety barrier against unplanned and unintended radiation exposures. The finding was of very low safety significance because the dose rates encountered and the workers short time period within the dose rate field were not sufficient to produce a substantial potential for an exposure in excess of regulatory limits.

To address this issue, the individuals involved were administratively precluded from entering the Radiologically Controlled Area for the remainder of the outage. Additionally, training to reinforce radiation protection standards and expectations was provided to radiation workers. One Non-Cited Violation for the failure to meet the requirements of Technical Specification 5.7.1.b regarding the control of activities in a high radiation area through a radiation work permit was identified. (Section 2OS1.5)

Green.

A finding of very low safety significance was self-revealed when a drum and trash bags of radioactive material were moved and created an unposted and unbarricaded high radiation area.

The issue was associated with the Human Performance and Program and Process attributes of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material because the uncontrolled high radiation area created the potential for unplanned and unintended dose to individuals working in the proximity of the drum and trash bags. The finding was of very low safety significance because the dose rates were not sufficient to produce a substantial potential for an exposure in excess of regulatory limits. Upon discovery, the licensee took immediate corrective actions to properly post the high radiation area. Additionally, further surveys were conducted to verify that no other unknown radiological conditions existed. One Non-

Cited Violation for the failure to meet the requirements of Technical Specification 5.7.1.a regarding barricading and posting a high radiation area was identified. (Section 2OS1.5)

B. Licensee Identified Findings None.

REPORT DETAILS

A list of documents reviewed within each inspection area is included at the end of the report.

Summary of Plant Status

The plant operated at full power during the inspection period with the following exception:

C On July 1, 2003, Main Turbine Stop Valve #2 inadvertently closed which resulted in a slight loss of load on the main generator and subsequent decrease in reactor power to 95 percent. Control room operators subsequently decreased reactor power to 87 percent. On July 6, 2003, after troubleshooting and necessary repairs were completed, control room operators re-opened Main Turbine Stop Valve #2. Reactor power was subsequently raised to full power on July 7,

REACTOR SAFETY

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

1R01 Adverse Weather Protection

.1 Inspection Scope

On July 7, 2003, control room operators received reports of severe thunderstorm warnings which included forecasted high wind gusts. The inspectors verified that prescribed actions in Off Normal Operating Procedure 12, Acts of Nature, were implemented as required for the predicted high wind conditions.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Quarterly Equipment Alignment Walkdowns

a. Inspection Scope

The inspectors performed three partial equipment alignment walkdowns of the following plant equipment:

C High Pressure Air System in the East Safeguards Room C Emergency Diesel Generator 1-2 C High Pressure Safety Injection Pump P-66A The inspectors performed the walkdowns to verify proper system lineup while redundant plant equipment was out of service. For the systems walked down, the inspectors verified that power was available, that accessible equipment and components were appropriately aligned, and that no discrepancies existed which would impact system function. Portions of the system alignment inspection included discussions and system walkdowns with operations and engineering personnel.

The inspectors also reviewed select condition reports related to equipment alignment issues and verified that identified problems were entered into the corrective action program with the appropriate significance characterization and that planned and completed corrective actions were appropriate.

b. Findings

No findings of significance were identified.

.2 Semiannual Equipment Alignment Walkdowns

a. Inspection Scope

The inspectors performed one complete walkdown inspection of the Critical Service Water System utilizing piping and instrumentation diagrams, system operating procedures, and system checklists to verify that accessible system components were correctly aligned. The inspectors also reviewed open maintenance work orders to verify that the equipments safety function was not adversely impacted.

The inspectors also reviewed select condition reports associated with the Critical Service Water System and verified that identified problems were entered into the corrective action program with the appropriate significance characterization and that planned and completed corrective actions were appropriate.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

The inspectors toured the following six areas in which a fire could affect safety-related equipment:

C West Safeguards Room (Fire Area 28)

C 1D Switchgear Room (Fire Area 3)

C Emergency Diesel Generator 1-2 Room (Fire Area 6)

C Battery #2 Room (Fire Area 11)

C Spent Fuel Pool Area (Fire Area 17)

C Safety Injection and Refueling Water Tank/Component Cooling Water Roof Area (Fire Area 32)

The inspectors assessed the material condition of the passive fire protection features and verified that transient combustibles and ignition sources were appropriately controlled. Also, the inspectors reviewed documentation for completed surveillances to verify the availability of the sprinkler fire suppression system, smoke detection system, and manual fire fighting equipment.

The inspectors verified that the installed fire protection equipment in the fire areas corresponded with the equipment which was referenced in the applicable portions of the Updated Final Safety Analysis Report, Section 9.6, Fire Protection.

The inspectors reviewed selected condition reports related to fire protection problems and verified that identified problems were entered into the corrective action program with the appropriate significance characterization and that planned and completed corrective actions were appropriate.

b. Findings

No findings of significance were identified.

1R06 Flood Protection

a. Inspection Scope

The inspectors performed one internal flood protection features inspection for the east engineered safeguards room which contained risk significant safety-related plant equipment.

The inspectors conducted walkdowns and design reviews, including reviews of preventive maintenance activities, for the following attributes associated with the room:

C Sealing of equipment below the floodline, such as electrical conduits; C Holes or unsealed penetrations in floors and walls between flood areas; C Adequacy of watertight doors between flood areas; and C Common drain system and sumps, including floor drain piping and check valves where credited for isolation of flood areas within plant buildings.

The inspectors also assessed condition reports related to flood protection issues to verify that identified problems were entered into the corrective action program with the appropriate significance characterization and that planned and completed corrective actions were appropriate.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

a. Inspection Scope

The inspectors observed one crew of reactor and senior reactor licensed operators during simulator training on September 17, 2003. The inspectors assessed the operators ability to use Off-Normal and Emergency Operating plant procedures to mitigate the following events:

C loss of main generator automatic voltage control; C sequential loss of three offsite power sources to the switchyard due to inclement weather with a subsequent loss of offsite power; C plant trip due to loss of offsite power concurrent with a failure of Emergency Diesel Generator 1-1; and C subsequent loss of Emergency Diesel Generator 1-2 resulting in a station blackout condition.

The inspectors also observed the post-scenario critique to assess the licensee evaluators and the crews ability to self-identify performance weaknesses.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

.1 Waste Gas Compressor C-50A Outage

a. Inspection Scope

The inspectors conducted one maintenance effectiveness inspection associated with the Waste Gas Compressor C-50A outage after a self-revealed tagging error resulted in work commencing prior to component cooling water to the compressor being appropriately isolated. The inspectors reviewed the activities and documentation associated with the work, including planning and scheduling; control room logs; and work order summaries. The inspectors assessed if the tagging error resulted in an adverse impact on the component cooling water system for which mitigating actions were required by the control room operators. The inspectors also reviewed other tagging orders for scheduled work on Compressor C-50A to determine if Administrative Procedure 4.10, Personnel Protective Tagging, requirements had been followed.

The inspectors searched corrective action documents to determine if there was an adverse trend related to inadequate tagging during maintenance activities. In addition, the inspectors reviewed Condition Report CAP036557, Incomplete Tagging Associated With Work on Waste Gas Compressor C-50A, to verify that the issue was entered into the corrective action program with the appropriate significance characterization.

b. Findings

No findings of significance were identified.

.2 Routine Maintenance Rule System Reviews

a. Inspection Scope

The inspectors conducted maintenance effectiveness inspections on the following two systems to assess the licensees maintenance rule program:

C Component Cooling Water System C Chemical Volume and Control System The inspectors reviewed the licensees maintenance rule performance indicators to verify that the system status had been appropriate categorized in accordance with the maintenance rule program. The inspectors reviewed work order histories and selected condition reports written against the system over the last 2 years to verify that maintenance and identified problems had been appropriately addressed. Completed work orders were reviewed to determine if there was an adverse trend in system performance that could be attributed to inappropriate work practices and to determine if there were any common cause issues that had not been addressed.

Further, the inspectors reviewed selected condition reports and associated maintenance rule evaluations to verify that identified problems were appropriately characterized and dispositioned in accordance with the licensees maintenance rule program. The inspectors also verified that planned corrective actions were appropriate and had been implemented as scheduled.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

The inspectors reviewed Operator Risk Reports, Shift Supervisor logs, and daily maintenance schedules to verify that equipment necessary to minimize plant risk was operable or available as required during planned and emergent maintenance activities.

The inspectors also conducted plant walkdowns to verify that equipment necessary to minimize risk was available for use. The following four activities were reviewed:

C Scheduled maintenance activities for Containment Spray Pump P-54B, High Pressure Air Compressor C-6B and Emergency Diesel Generator 1-1 Testing concurrent with emergent work activities associated with Emergency Diesel Generator 1-1 on July 21-25, 2003; C Scheduled maintenance on Emergency Diesel Generator 1-2 on August 6-7, 2003; C Emergent activities associated with High Pressure Safety Injection Pump P-66B Subcooling Control Valve CV-3070 and the loss of the electrical transmission grid that occurred on the East Coast during the week of August 11, 2003; and C Scheduled maintenance activities on Emergency Diesel Generator 1-1 on September 16-17, 2003.

b. Findings

No findings of significance were identified.

1R14 Operator Performance During Non-Routine Evolutions and Events

.1 Operator Response to Loss of Load

a. Inspection Scope

The inspectors observed operator response to an unexpected closure of Turbine Stop Valve #2 on July 1, 2003. The inspectors also verified that the actions prescribed in Off Normal Procedure 1, Loss of Load, were appropriately implemented.

b. Findings

No findings of significance were identified.

.2 Operator Response to Automatic Start of Both Emergency Diesel Generators Due to the

Loss of Grid on the East Coast

a. Inspection Scope

On August 14, 2003, the inspectors observed the operator response to the automatic start of Emergency Diesel Generators 1-1 and 1-2 which resulted from the voltage drop on the 2140 Volt safety busses due to the loss of the electrical transmission grid on portions of the east coast of the United States and parts of Canada. The inspectors walked down the control panels to verify that plant equipment responded as designed and that the off site power sources to the plant switchyard remained available.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed four operability assessments as documented in the associated condition reports for the following risk significant plant equipment:

• Service Water System;

• Containment Sump Check Valve CK-ES3181;

• Containment Air Cooler Fan V-4A; and,

• Charging Pump P-55A.

The inspectors interviewed the cognizant engineers and reviewed the supporting documents to assess the adequacy of the operability assessments for the current plant mode. The inspectors also reviewed the applicable sections of the Technical Specifications, Updated Final Safety Analysis Report, and Design Basis Documents to verify that the operability assessments were technically adequate and that the components remained available, such that no unrecognized increase in plant risk had occurred.

b. Findings

The inspectors identified one finding of very low safety significance (Green) pertaining to Containment Air Cooler Fan V-4A.

Introduction The inspectors determined that a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed when Containment Air Cooler Fan V-4A tripped unexpectedly on July 1, 2003, after licensee personnel declared the fan operable on June 20, 2003, following emergent repairs.

Description On June 19, 2003, during a routine containment entry, an Auxiliary Operator discovered loose parts which included nuts, bolts, and a washer, under safety-related Containment Air Cooler Fan V-4A and noted elevated noise levels from the fan. Licensee personnel subsequently determined that the loose parts were fasteners from the fan ductwork and the noise was from the loose ductwork. Consequently, Fan V-4A was declared inoperable due to concerns regarding the integrity of the ductwork during a seismic event.

The licensee reinstalled the loose fasteners on June 20, 2003, to re-establish the ductwork integrity. Following these repairs, licensee personnel obtained Fan V-4A motor current readings which were found to be higher than previous readings, but considered acceptable. Vibration data was also obtained from the fan housing and was considered high when compared to general industry standards, however no comparable baseline vibration data existed for this fan. Licensee personnel also manually rotated the fan and noted that the fan would not rotate without the use of continuous manual force. Consequently, licensee personnel suspected bearing degradation in the fan motor, but concluded that bearing failure was not imminent.

Fan V-4A was subsequently declared operable and returned to service on June 20, 2003. The operability determination was based on information obtained from the visual inspections, the fan motor current readings, and the vibration data from which licensee personnel concluded that Fan V-4A could be returned to service and imminent failure would not occur. However, on July 1, 2003, Containment Air Cooler Fan V-4A tripped unexpectedly due to a failed motor drive end bearing and Fan V-4A was again declared inoperable.

Technical Specification 3.6.6, Containment Cooling Systems, Condition A was entered which required that Fan V-4A be returned to an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

However, because planned repairs for Fan V-4A required more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, enforcement discretion to complete the repairs with the plant on-line and avoid a plant shutdown was requested by licensee personnel and granted by the NRC on July 3. The Notice of Enforcement Discretion is discussed in Section 4OA5 of this report. Licensee personnel subsequently completed the necessary repairs and Fan V-4A was declared operable on July 6, 2003.

The inspectors noted that when the fan was declared operable on June 20, 2003, that there was no formal operability recommendation form completed, which would have required a more technically rigorous operability evaluation than was performed.

Therefore, the inspectors concluded that the lack of rigor in the technical evaluation to determine operability for Fan V-4A on June 20 led to a non-conservative operability determination. Consequently, Fan V-4A was declared operable and returned to service with more significant motor bearing degradation than was recognized by licensee personnel which rendered the fan incapable of performing the required safety function of containment atmosphere air mixing for 30 days following a design basis accident.

Analysis The inspectors determined that the lack of rigor in the operability determination completed on June 20 for Fan V-4A was a licensee performance deficiency warranting a significance evaluation. The Barrier Integrity cornerstone was impacted by this issue.

The inspectors reviewed the samples of minor issues in Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues and determined that there were no examples that appropriately described this issue. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, Issue Disposition Screening, because it was related to the Human Performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that the containment physical design barrier protects the public from radionuclide releases caused by accidents because Fan V-4A was unable to perform its associated safety function when returned to service on June 20. This finding was also associated with the cross-cutting area of Problem Identification and Resolution which is briefly discussed in Section 4OA4 of this report.

The inspectors determined that the finding could be evaluated using IMC 0609, Significance Determination Process, (SDP) because the finding was associated with the integrity of reactor containment. Using IMC 0609, Appendix A, SDP Phase 1 Screening Worksheet for IE [Initiating Events], MS [Mitigating Systems], and B [Barrier Integrity] Cornerstones, the inspectors determined that the Barrier Integrity cornerstone was the only affected area. Using only the Barrier Integrity column on the worksheet, the inspectors determined that the finding

(1) did not represent only a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool;

(2) did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere; and

(3) did not represent an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment. Therefore, the finding screened out as Green and was considered to be of very low safety significance.

Enforcement 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to this, the extent of motor bearing degradation on safety-related Containment Air Cooler Fan V-4A, a condition adverse to quality, was not identified and promptly corrected after the condition was initially discovered on June 19, 2003. Consequently, Fan V-4A was declared operable and returned to service on June 20, 2003, and subsequently tripped on July 1, 2003, because the motor bearing failed. Therefore, Fan V-4A would not have been able to perform its safety function to ensure proper mixing of the containment atmosphere following a design basis accident for 30 days.

However, because this violation was associated with a finding of very low safety significance and because the finding was entered into the licensees corrective action program, this violation is being treated as an Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000255/2003006-01).

The licensee entered the issue into the corrective action program as CAP036444 and CAP036565. A root cause evaluation was also completed. Corrective actions to address the issue included replacing the motor for Fan V-4A and entering all containment air cooler fans and motors into a predictive maintenance program.

1R19 Post Maintenance Testing

a. Inspection Scope

The inspectors observed portions of post maintenance testing and reviewed documented testing activities to verify that the tests were adequately performed for the following seven activities:

C Containment Air Cooler Ventilation Fan V-4A C Auxiliary Feedwater Pump P-8A Breaker 152-104 Replacement C Auxiliary Feedwater to A Steam Generator Control Valve CV-0749 C Emergency Diesel Generator 1-2 C Containment Spray Pump P-54C Breaker 152-114 Replacement C High Pressure Safety Injection Pump P-66B Subcooling Control Valve CV-307 C Emergency Diesel Generator 1-1.

The inspectors verified that applicable testing prerequisites were met prior to the start of the tests and that the effect of testing on plant conditions was adequately addressed by the control room operators.

The inspectors also reviewed

(1) post maintenance testing criteria to verify that the test criteria and acceptance criteria were appropriate for the scope of work performed;

(2) completed tests and associated procedures to verify that the tests adequately verified system operability; and

(3) documented test data to verify that the data was complete and that the equipment met the testing acceptance criteria.

The inspectors also reviewed condition reports to verify that post maintenance testing problems were entered into the corrective action process with the appropriate significance characterization.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the following four surveillance testing activities conducted on risk-significant plant equipment:

C ED-01 and ED-02 Station Battery Checks C Auxiliary Feedwater Automatic Initiation C Emergency Diesel Generator 1-2 Load Rejection C ATWS (Anticipated Transient Without Scram) Functional Testing The inspectors observed portions of the testing in the plant to verify that the testing was conducted in accordance with prescribed procedures. The inspectors also reviewed the documented test data for the Technical Specification Surveillance Test procedures and the associated basis documents to verify that testing acceptance criteria were satisfied.

In addition, the inspectors reviewed applicable portions of Technical Specifications, the Updated Final Safety Analysis Report and Design Basis Documents to verify that the surveillance tests adequately demonstrated that system components could perform required safety functions.

Further, the inspectors reviewed selected condition reports regarding surveillance testing activities to verify that the identified problems were entered into the licensees corrective action program with the appropriate significance characterization and that planned and completed corrective actions were appropriate.

b. Findings

The inspectors identified one Unresolved Item pertaining to ATWS system testing.

Introduction The inspectors identified an Unresolved Item related to the failure to functionally test the ATWS system circuitry which provided an automatic start signal to Turbine Driven Auxiliary Feedwater Pump P-8B.

Description In 1990, the licensee completed a plant modification which installed ATWS equipment as described in Updated Final Safety Analysis Report, Section 7.2, to satisfy the 10 CFR 50.62 ATWS rule. Included in the modification was circuitry to provide an automatic start signal to Turbine Driven Auxiliary Feedwater Pump P-8B on a loss of direct current (DC) control power.

While reviewing surveillance testing for the Auxiliary Feedwater System, the inspectors questioned licensee personnel regarding which procedure tested the ATWS system function to automatically start Pump P-8B and when testing was last completed.

Licensee personnel subsequently determined that the function had not been tested since the 1999 refueling outage and generated Condition Report CAP036974, Failure to Perform ATWS Steam Driven Aux Feedwater Pump Test RPS-I-10, which was entered into the corrective action program and required a condition evaluation.

Through the evaluation, licensee personnel determined that in 1991, a commitment was made to the NRC to implement periodic surveillance testing of the ATWS system and to implement end to end functional testing of the system during refueling outages. The NRC opened Unresolved Item (50-255/91002-01(DRS)) pending review of the licensees proposed ATWS system testing. The licensees commitment and the associated unresolved item were documented in Inspection Report 50-255/91002(DRS).

Licensee personnel subsequently developed and implemented ATWS system test procedures in 1991 and Unresolved Item 50-255/91002-01(DRS) was closed as documented in Inspection Report 50-255/94004. From the time that the ATWS system test procedures were developed through the 1999 refueling outage, the ATWS system function to automatically start Turbine Driven Auxiliary Feedwater Pump P-8B tested satisfactorily in accordance with test procedure RPS-I-8, Anticipated Transient Without Scram (ATWS)/PORV [Power Operated Relief Valve] High Pressurizer Pressure Actuation Functional Test.

During the 1999 refueling outage, testing activities for the ATWS system function to automatically start Pump P-8B were delayed because plant conditions would not support testing. Similar delays in testing had also occurred during previous outages. As a result, after the 1999 refueling outage, licensee planning and maintenance personnel determined that it would be more efficient to test the ATWS function to automatically start Pump P-8B in a separate procedure instead of testing that function within procedure RPS-I-8. Therefore, Test Procedure RPS-I-10, Aux Feed Pump K8 Auto Start on Loss of AFAS [Auxiliary Feedwater Actuation Signal] DC Control Power, was developed and issued on May 24, 2000.

However, licensee personnel failed to generate a preventative maintenance activity that scheduled RPS-I-10 during subsequent refueling outages. Consequently, the ATWS system function to automatically start Pump P-8B was not tested during the 2001 and 2003 refueling outages as committed to by the licensee in 1991 and discussed in the Updated Final Safety Analysis Report. Although testing the ATWS system function to automatically start Pump P-8B had not been completed since the 1999 refueling outage, the inspectors reviewed past testing and determined that the automatic start function of Pump P-8B had been completed satisfactorily on all occasions prior to 1999.

During the condition report evaluation, licensee personnel developed the following corrective actions to address this finding:

C Preventive maintenance activity PPAC RPS-023, Performance of RPS-I-10, was developed to schedule surveillance test RPS-I-10 every refueling outage C Work Request 296123 was initiated to perform surveillance test RPS-I-10 during the next forced outage of sufficient duration should one occur before the next refueling outage.

The inspectors verified that the preventative maintenance activity and the work request were entered into the licensees work management system.

This is an Unresolved Item (URI 05000255/2003006-02) pending a review of the ATWS system testing results.

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the modification documentation and the associated 10 CFR 50.59 evaluation for temporary plant modification TM-2003-024, Open Links to Safety Injection Tank Pressure Control Solenoid Valves.

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 Fleet Modification Procedure FP-E-MOD-03, Temporary Modifications. The inspectors verified that the temporary modification was implemented in the plant as designed, appropriately controlled, and that required plant drawing and procedure revisions were completed. The inspectors also reviewed post-installation test results to verify that testing was completed satisfactorily and that the impact of the temporary modification on the safety injection tank pressure control valves was adequately evaluated.

In addition, the inspectors reviewed condition reports to verify that temporary modification problems were entered into the corrective action program with the appropriate significance characterization.

b. Findings

No findings of significance were identified.

1EP6 Emergency Preparedness Drill Evaluation

a. Inspection Scope

The inspectors observed activities in the plant simulator, Technical Support Center and the Emergency Offsite Facility during an emergency preparedness drill conducted on September 24, 2003. The inspectors verified that the emergency classifications, notifications to offsite agencies, and protective action recommendations were completed in an accurate and timely manner as required by the emergency plan implementing procedures. The inspectors also verified that the drill was conducted in accordance with the prescribed sequence of events and that the drill objectives were met.

The inspectors observed the post-drill critique in the Technical Support Center to verify that licensee personnel and licensee drill evaluators adequately self-identified drill performance problems. The inspectors also verified that condition reports concerning drill performance problems were generated and entered into the corrective action program with the appropriate significance characterization.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas (71121.01)

.1 Review of Licensee Performance Indicators for the Occupational Exposure Cornerstone

a. Inspection Scope

The inspectors reviewed the licensees occupational exposure control cornerstone performance indicators (PIs) to determine whether or not the conditions surrounding the PIs had been evaluated, and identified problems had been entered into the corrective action program for resolution.

This review represented one inspection sample.

b. Findings

No findings of significance were identified.

.2 Plant Walkdowns and Radiation Work Permit (RWP) Reviews

a. Inspection Scope

The inspectors assessed the adequacy of the licensees internal dose assessment process for two internal exposures of greater than or equal to 50 millirem committed effective dose equivalent for workers involved in the In-Core Instrumentation work activities during the most recent refueling outage.

The inspectors also reviewed the licensees physical and programmatic controls for highly activated and/or contaminated materials (non-fuel) stored within spent fuel or other storage pools.

These reviews represented two inspection samples; one sample for the review of the adequacy of the licensees internal dose assessment process and one sample for the review of the licensees controls of stored radioactive material.

b. Findings

No findings of significance were identified.

.3 Problem Identification and Resolution

a. Inspection Scope

The inspectors reviewed seven corrective action reports related to access controls and two high radiation area radiological incidents. Staff members were interviewed and corrective action documents were reviewed to verify that follow-up activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk based on the following:

Initial problem identification, characterization, and tracking

Disposition of operability/reportability issues

Evaluation of safety significance/risk and priority for resolution

Identification of repetitive problems

Identification of contributing causes

Identification and implementation of effective corrective actions.

Resolution of Non-Cited Violations (NCVs) tracked in the corrective action system

Implementation/consideration of risk significant operational experience feedback The inspectors evaluated the licensees process for problem identification, characterization, and prioritization, and verified that problems were entered into the corrective action program and resolved. For repetitive deficiencies and/or significant individual deficiencies in problem identification and resolution, the inspectors verified that the licensees self-assessment activities were capable of identifying and addressing these deficiencies.

These reviews represented two inspection samples; one sample for the review of access control issues and one sample for the review of high radiation area incidents.

b. Findings

No findings of significance were identified.

.4 Job-In-Progress Reviews

a. Inspection Scope

Radiological work in high radiation work areas having significant dose rate gradients was reviewed to evaluate the application of dosimetry to effectively monitor exposure to personnel and to verify that licensee controls were adequate. Specifically, the inspectors reviewed the licensees enhanced exposure controls for the non-destructive evaluation of the bare metal reactor vessel head and the steam generator nozzle dam installation and removal work activities conducted during the most recent refueling outage. These work areas involved areas with significant dose rate gradients which increased the necessity of providing multiple dosimeters and/or enhanced job controls.

This represented one inspection sample for the review of radiological work in high radiation work areas having significant dose rate gradients.

b. Findings

No findings of significance were identified.

.5 Radiation Worker Performance and Radiation Protection Technician Proficiency

a. Inspection Scope

The inspectors reviewed condition reports generated during or since the previous refueling outage which identified that the root cause of the event was related to radiation worker errors or radiation protection technician errors to determine if there was a trend due to a similar cause, and to determine if this perspective matched the corrective action approach taken by the licensee to resolve the reported problems. These problems, along with planned and accomplished corrective actions, were discussed with the Radiation Protection Manager.

These reviews represented two inspection samples; one sample for the reviews related to radiation worker errors, and one sample for the reviews related to radiation protection technician errors.

b. Findings

Introduction Three self-revealed Green findings and associated Non-Cited Violations (NCVs) were identified when, during the most recent refueling outage,

(1) two workers entered a High Radiation Area without obtaining a briefing regarding the radiological conditions in the area;

(2) one of the two workers failed to stop work and report to the Radiation Protection Department when an electronic dosimetry dose rate alarm was received; and

(3) radioactive material was moved and created an unposted and unbarricaded High Radiation Area.

Description During the most recent refueling outage, on April 15, 2003, a Containment Area Coordinator (CAC) on the refueling floor (690 foot elevation) was assigned to move trash bags and a drum out of a posted High Radiation Area (HRA). The drum, which was labeled as radioactive material with dose rate information, contained contaminated stud hole plugs and guide pins previously used in the reactor cavity area. The CAC was to contact and obtain RP support for the evolution prior to commencing work. The CAC contacted an RP technician (RPT) on the refueling floor. According to the RPT account after the event, the technician indicated that he was busy and could not support the evolution at the time, but he would contact the CAC later to assist in the evolution.

However, the CAC believed that the conversation between the RPT and himself was satisfactory and that he had permission to proceed with the evolution. Shortly thereafter, the CAC obtained the services of a contract worker, and directed the worker to enter the posted HRA where the trash and drum were stored. However, there were no RP personnel in the general area to provide work coverage, nor were any briefings provided to the CAC and worker regarding radiological conditions, expected dose rates, or electronic dosimetry (ED) alarm settings prior to their entry.

The worker encountered difficulties while attempting to place a sling around the drum in preparation for moving the drum outside of the HRA. Subsequently, the worker received an ED dose rate alarm while attempting to move the drum. During the licensees investigation, the worker indicated that he heard the ED alarm, but did not know what actions were required upon receipt of the dose rate alarm. The worker did not inform the CAC that he had received the alarm nor were any RP personnel immediately contacted about the alarm. Rather, the worker continued working and moved the drum outside of the posted HRA. About 10 minutes later, the CAC moved four bags of trash, which were also labeled as radioactive material with dose rate information, from the posted HRA and placed them in a box adjacent to the drum of radioactive material.

During the CACs movement of the radioactive material trash bags, he also received an ED dose rate alarm, however, during the licensees investigation of the event, the CAC stated that he did not hear a dose rate alarm while he was moving the trash.

When both workers later exited the radiologically controlled area (RCA) and attempted to log out their EDs, they both received messages to contact RP because they had received dose rate alarms during their entries. According to the licensees investigation, during initial interviews with the CAC and worker, it became apparent that the workers received their ED dose rate alarms while working in a posted HRA and their subsequent actions of moving the drum and trash resulted in the creation of a new, unposted and unbarricaded HRA on the refuel floor for a period of about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The RP staff subsequently took actions to survey, barricade, and post the area around the drum and trash as a HRA, and the two workers were administratively locked out of the RCA for the remainder of the refueling outage.

Analysis

(1) The inspectors determined that the workers failure to obtain a radiological briefing prior to their entry into the HRA was a performance deficiency warranting a significance evaluation. The Occupational Radiation Safety cornerstone was impacted by this issue.

The inspectors reviewed the samples of minor issues in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, and determined that there were no examples similar to this issue. The inspectors concluded that the finding was of more than minor risk significance in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, since the finding was associated with the Human Performance attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material because the workers were not sufficiently cognizant of the radiation fields they could have encountered while inside the HRA and the issue involved the unplanned and unintended dose, or the potential of such a dose, resulting from actions contrary to Technical Specifications and licensee procedures.

Utilizing IMC 0609, Significance Determination Process, Appendix C, Occupational Radiation Safety SDP, the inspectors determined that the finding

(1) did not involve ALARA/work controls,

(2) did not result in an overexposure, and

(3) based on the surveys of the material inside the HRA and length of time the workers spent in the HRA, did not result in a substantial potential for an overexposure or compromise the licensees ability to assess dose. Consequently, the finding screened out as Green and was of very low safety significance.

Enforcement

(1) Technical Specification 5.7.1.e requires, in part, that an entry into a High Radiation Area be made only after dose rates in the area have been determined and personnel entering the area are knowledgeable of these dose rates. Contrary to the above, on April 15, 2003, two workers failed to obtain a radiological briefing and become knowledgeable of the dose rates prior to their entry into a High Radiation Area which was a violation of Technical Specification 5.7.1.e. However, because this violation was associated with a finding of very low safety significance and because the finding was entered into the licensees corrective action program, this violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000255/2003006-03). This violation was entered into the licensees corrective action program as CAP035210/RCE000330. To address this issue, the individuals involved were administratively precluded from entering the Radiologically Controlled Area for the remainder of the outage. Additionally, training to reinforce radiation protection standards and expectations was provided to radiation workers.

Analysis

(2) The inspectors determined that the failure of the worker to stop work and contact RP upon receiving an ED alarm was a performance deficiency warranting a significance evaluation. The Occupational Radiation Safety cornerstone was impacted by this issue.

The inspectors reviewed the samples of minor issues in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, and determined that there were no examples similar to this issue. The inspectors concluded that the finding was of more than minor risk significance in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, since the finding was associated with the Human Performance attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material because the issue involved a workers unplanned and unintended dose, or the potential of such a dose, resulting from actions contrary to licensee Technical Specifications and RWP requirements.

The inspectors determined that the workers failure to stop work and contact RP upon receiving an ED dose rate alarm was a failure of the individual radiation safety barrier against unplanned and unintended radiation exposures. Additionally, all RWPs at Palisades contain the actions workers shall take upon receipt of ED dose and dose rate alarms. Further, both workers attended the stations Back to Basics training prior to the outage which emphasized proper radiation worker practices, including the proper response to ED alarms.

Utilizing IMC 0609, Significance Determination Process, Appendix C, Occupational Radiation Safety SDP, the inspectors determined that the finding

(1) did not involve ALARA/work controls,

(2) did not result in an overexposure, and

(3) based on the surveys of the material inside the HRA and length of time the workers spent in the HRA, did not result in a substantial potential for an overexposure or compromise the licensees ability to assess dose. Consequently, the finding screened out as Green and was of very low safety significance.

Enforcement

(2) Technical Specification 5.7.1.b requires that the access to and activities in a High Radiation Area shall be controlled by means of a Radiation Work Permit, or equivalent, that includes the radiation dose rates in the work area and other requirements regarding necessary radiation protection equipment and measures. Radiation Work Permit P03-5100, which controlled activities conducted on the refueling floor on April 15, 2003, required that in the event of an electronic dosimetry dose rate alarm, the worker back out of the area, contact a Radiation Protection Technician, and await further instructions.

Contrary to the above, on April 15, 2003, during activities on the refueling floor, a worker failed to back out of an area, contact a Radiation Protection Technician, and await further instructions upon receiving an electronic dosimetry dose rate alarm which was not in accordance with Radiation Work Permit P03-5100 and was a violation of Technical Specification 5.7.1.b. However, because this violation was associated with a finding of very low safety significance and because the finding was entered into the licensees corrective action program, this violation is being treated as a Non-Cited Violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000255/2003006-04). This violation was entered into the licensees corrective action program as CAP035210/RCE000330. To address this issue, the individuals involved were administratively precluded from entering the Radiologically Controlled Area for the remainder of the outage. Additionally, training to reinforce radiation protection standards and expectations was provided to radiation workers.

Analysis

(3) The inspectors determined that the movement of the drum and trash which resulted in the creation of an unposted and unbarricaded HRA for about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> was a performance deficiency warranting a significance evaluation. The Occupational Radiation Safety cornerstone was impacted by this issue. The inspectors reviewed the samples of minor issues in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, and determined that there were no examples similar to this issue. The inspectors concluded that the finding was of more than minor risk significance in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, since the finding was associated with the Human Performance and Program and Processes attributes of the Occupational Radiation Safety cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material because the issue involved the occurrence of the potential for unplanned, unintended dose to other individuals working near the unposted, unbarricaded HRA resulting from actions contrary to licensee Technical Specifications.

Utilizing IMC 0609, Significance Determination Process, Appendix C, Occupational Radiation Safety SDP, the inspectors determined that the finding

(1) did not involve ALARA/work controls,

(2) was not associated with an overexposure, and

(3) based on the surveys of the radioactive drum and trash, did not result in a substantial potential for an overexposure or compromise the licensees ability to assess dose. Consequently, the finding screened out as Green and was of very low safety significance.

Enforcement

(3) Technical Specification 5.7.1.a requires, in part, that each entryway to a High Radiation Area shall be barricaded and conspicuously posted as a High Radiation Area. Contrary to the above, on April 15, 2003, radioactive material consisting of a drum and trash bags relocated from a posted High Radiation Area on the refueling floor to another location on the refueling floor created a High Radiation Area which was not posted and barricaded and was a violation of Technical Specification 5.7.1.a. However, because this violation was associated with a finding of very low safety significance and because the finding was entered into the licensees corrective action program, this violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000255/2003006-05). This violation was entered into the licensees corrective action program as CAP035210/RCE000330. Upon discovery, the licensee took immediate corrective actions to properly post the high radiation area. Additionally, further surveys were conducted to verify that no other unknown radiological conditions existed.

2OS2 As-Low-As-Is-Reasonably-Achievable (ALARA) Planning And Controls (71121.02)

.1 Inspection Planning

a. Inspection Scope

The inspectors reviewed plant collective exposure history, current exposure trends, and ongoing and planned activities in order to assess current performance and exposure challenges. This included determining the plants current 3-year rolling average for collective exposure in order to help establish resource allocations and to provide a perspective of significance for any resulting inspection finding assessment. The inspectors determined site specific trends in collective exposures and source-term measurements.

These reviews represented one inspection sample for the review of collective radiation exposure for the previous 3 years.

b. Findings

No findings of significance were identified.

.2 Radiological Work Planning.

a. Inspection Scope

The inspectors compared the results achieved including dose rate reductions and person-rem used with the intended dose established in the licensees ALARA planning for planned work activities. Reasons for inconsistencies between intended and actual work activity doses were reviewed. The inspectors reviewed the RWP/ALARA reviews for the following seven work activities from the most recent refueling outage:

• Reactor Head Disassembly and Movement (RWP P03-5102)

• Upper Guide Structure Lift Rig/In-core Instrumentation Activities (RWP P03-5104)

• Reactor Head Reassembly/Closeout Activities (RWP P03-5108)

• In-Core Instrumentation Flange Activities (RWP P03-5111)

• Nozzle Dam Installation/Removal Activities (RWP P03-5150)

• ROSA [Remotely Operated Service Arm]/Eddy Current Testing and Tube Plugging (RWP P03-5152)

• Containment Scaffold Work (RWP P03-5306)

The inspectors compared the person-hour estimates, provided by maintenance planning and other groups, with the actual work activity time requirements in order to evaluate the accuracy of these time estimates. The licensees post-job (work activity) reviews were evaluated to verify that identified problems were properly entered into the licensees corrective action program.

These reviews represented three inspection samples; one sample for the review of ALARA planning, one sample for the review of person-hour estimates, and one sample for the review of problems entered into the corrective action program.

b. Findings

No findings of significance were identified.

.3 Problem Identification and Resolutions

a. Inspection Scope

The inspectors reviewed the licensees ALARA program self-assessments since the last inspection to determine if the licensees overall audit programs scope and frequency met the requirements of 10 CFR 20.1101(c).

The inspectors verified that identified problems were entered into the corrective action program for resolution, and that they had been properly characterized, prioritized, and resolved. This included dose significant post-job (work activity) reviews and post-outage ALARA report critiques of exposure performance.

Corrective action reports related to the ALARA program were reviewed and staff members were interviewed to verify that follow-up activities had been conducted in an effective and timely manner commensurate with their importance to safety and risk using the following criteria:

Initial problem identification, characterization, and tracking

Disposition of operability/reportability issues

Evaluation of safety significance/risk and priority for resolution

Identification of repetitive problems

Identification of contributing causes

Identification and implementation of effective corrective actions

Resolution of Non-Cited Violations tracked in the corrective action system

Implementation/consideration of risk significant operational experience feedback These reviews represented three inspection samples; one sample for the review of ALARA program self-assessments, one sample for the review of problems entered in the licensees corrective action program, and one sample for the review of follow-up activities related to corrective action reports.

b. Findings

No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03)

.1 Respiratory Protective Equipment Maintenance and User Training

a. Inspection Scope

The inspectors reviewed the licensee's respiratory protection and confined space entry procedures and discussed their implementation relative to the requirements of 10 CFR 20.1703(f) for standby rescue persons whenever one-piece atmosphere supplying suits, or any combination of respiratory protection and personnel protective equipment were used which the wearer may have difficulty extricating himself.

Specifically, the inspectors reviewed the licensee's work planning process and implementing practices, and interviewed RP staff regarding the following aspects of 10 CFR 20.1703:

(1) designation of an adequate number of standby rescue workers and their training/instruction,

(2) presence of equipment staged at the work site for the safety of the rescuer and for extrication of the respiratory equipment user,

(3) practices for continuous communication between standby rescuer(s) and the respiratory protection user(s), and

(4) provisions for immediate availability of the standby rescuer.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

.1 Reactor Safety Performance Indicators

a. Inspection Scope

The inspectors reviewed the data submitted by licensee personnel for July 2002 through June 2003 to verify that the following two Performance Indicators were reported accurately:

• Heat Removal System (Auxiliary Feedwater System) Unavailability

• Residual Heat Removal System Unavailability The inspectors reviewed samples of records regarding maintenance rule performance, control room logs, maintenance activities which resulted in unavailability time, and monthly operating data reports.

b. Findings

No findings of significance were identified.

.2 Radiation Safety Strategic Area

a Inspection Scope The inspectors sampled licensee submittals for the performance indicators (PI) listed below for the period from October 2002 to June 2003. To verify the accuracy of the PI data reported during that period, PI definitions and guidance contained in Revision 2 of Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, were used.

• Occupational Exposure Control Effectiveness The inspectors previously reviewed the one unintended exposure occurrence under this PI which occurred in November 2002 and was documented in NRC Inspection Report 05000255/2003002. Since no additional reportable events were identified by the licensee for the 4th quarter of calendar year 2002 through the 2nd quarter of calendar year 2003, the inspectors compared the licensees data with the corrective action program database and the radiological controlled area exit electronic dosimetry transaction records for these time periods to verify that there were no unaccounted for occurrences in the Occupational Radiation Safety Performance Indicator. Additionally, the inspectors conducted walkdowns of accessible locked high radiation areas and very high radiation area entrances to verify the adequacy of controls in place for these areas.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Loss of Bus 1E Due to Removal of Start-up Transformer 1-2 Undervoltage Potential

Transformer Fuses

a. Inspection Scope

On April 1, 2003, power was lost to nonsafety-related 2400 Volt bus 1E when incorrect potential transformer fuses were removed from safety-related 2400 Volt bus 1D. This resulted in the interruption of the plant service air supply to steam generator nozzle dams during a period of reduced primary coolant system inventory. The inspectors previously documented a finding of very low safety significance and a violation of Technical Specification 5.4.1 (NCV 50-255/03-04-01) related to this issue. The inspectors reviewed the licensees root cause evaluation for the following condition report associated with this event:

C CAP034788, Loss of Bus 1E Due to Removal of Start-up Transformer 1-2 Undervoltage Potential Transformer Fuses" The inspectors verified the following attributes during their review of the licensee's root cause evaluation and corrective actions:

C evaluation and disposition of performance issues and operability issues; C consideration of the extent of condition, generic implications, common cause and previous occurrences; C classification and prioritization of the resolution of the problem, commensurate with safety significance; C identification of the root and contributing causes of the problem; and C identification of corrective actions which were appropriately focused to correct the problem.

The inspectors discussed the corrective actions and condition report evaluation with site personnel.

b. Findings and Observations

No findings of significance were identified. The inspectors verified that the root cause evaluation and associated corrective actions were appropriate. However, the inspectors noted one minor weakness in the evaluation.

The licensee concluded that the root cause for this event was that maintenance personnel failed to meet station standards for procedure use and adherence. The inspectors noted that the root cause evaluation was not as critical of the role that operators had in causing the event. Operators did not participate in the pre-job brief for the maintenance activity, which included the removal of potential transformer fuses to de-energize the metering circuitry.

The evaluation stated that operators role was limited and straightforward and that their participation in the pre-job brief would have been beneficial only if the fuse identity confusion had been known in advance. The evaluation also stated that the ability to coordinate their attendance would have been restricted by the reduced amount of time available to plan and perform the work. As a result, operators were not familiar with the work order and operators did not use the work order to confirm that the actions taken were correct before removing the fuses.

The inspectors concluded that had operators been involved in the pre-job brief and reviewed the work order, they would likely have recognized that only one set of fuses were to be removed, which could have precluded this event. However, while the documented evaluation was not as critical to the role the operators had in causing the event, the inspectors determined that the identified corrective actions for this issue were adequate and that they also addressed the associated human performance deficiencies demonstrated by the operators.

.2 Diluted Boric Acid in the Chemical Volume Control System Blender Line

a. Inspection Scope

On May 8, 2003, an unexpected increase in reactor power and primary coolant system temperature occurred following a routine blend to the Volume Control Tank. Control room operators subsequently inserted control rods and reduced load on the main turbine to mitigate the unexpected response and to ensure that steady state reactor thermal power limits were not exceeded. The inspectors reviewed the Apparent Cause Evaluation for Condition Report CAP035633, Did Not See the Effects of Boron During Blend to the Volume Control Tank, that was generated for this issue. The inspectors verified that the problem was accurately identified; the apparent cause was adequately justified; extent of condition and generic implications were appropriately addressed; and that corrective actions were appropriately focused to address the problem and implemented commensurate with the safety significance of the issue.

b. Findings

No findings of significance were identified. The inspectors determined the identified cause was appropriately justified and that the identified corrective actions had been implemented or were scheduled to be implemented commensurate with the safety significance of the issue. However, the inspectors noted one minor weakness regarding problem identification.

During the apparent cause evaluation, licensee personnel determined that two valves not associated with the apparent cause in the Chemical Volume Control System may have been leaking by causing a minor amount of dilution in the boric acid pumped feed line. Based on data taken during the evaluation, licensee personnel determined that the dilution was either from the primary coolant system water through Check Valve CK-CVC2141 or from the primary makeup water system water through Manual Valve MV-CVC2167. However, the inspectors noted that no condition report or work request had been generated to ensure that the identified valve deficiencies would be addressed.

The inspectors concluded that the amount of dilution in the boric acid pumped feed line would not result in any adverse consequences of significance if the primary coolant system was borated using the pumped feed line. Consequently, this issue was considered minor; however, the failure to generate a condition report or work request regarding the potentially leaking valves demonstrated a weakness in entering identified problems into the corrective action program in a timely manner.

Licensee personnel subsequently generated Condition Report CAP037950, Leakage Into Boric Acid Pumped Feed Line Identified, which was entered into the licensees corrective action program to evaluate the identified condition and develop corrective actions as necessary. The inspectors verified that this issue was entered into the corrective action program with the appropriate significance characterization.

.3 Failure of Containment Fan Cooler V-4A

a. Inspection Scope

The inspectors reviewed the root cause evaluation associated with Condition Report CAP036444, Containment Air Cooler Fan V-4A Tripped Unexpectedly. The inspectors verified that the identification of the problem was complete, accurate and identified in a timely manner commensurate with its ease of discovery; that the evaluation and disposition of performance issues and operability issues was adequate; the root cause was adequately justified; extent of condition and generic implications were appropriately addressed; and that corrective actions were appropriately focused to address the problem and implemented commensurate with the safety significance.

b. Findings

No findings of significance were identified. The inspectors determined that the identified root cause was appropriately justified and that the identified corrective actions were adequate and had been implemented or were scheduled to be implemented commensurate with the safety significance of the issue. However, the inspectors noted one weakness regarding problem identification which had been documented in the evaluation.

On June 19, 2003, an Auxiliary Operator discovered loose parts consisting of nuts and bolts under Containment Air Cooler Fan V-4A and noted an abnormal noise coming from the fan motor. Licensee personnel subsequently determined that the loose parts were fasteners that had come from the associated ductwork and the noise was from the loose ductwork. Consequently, Fan V-4A was declared inoperable due to integrity concerns with the ductwork during a seismic event. Fan V-4A was declared operable and returned to service on June 20, 2003, after the fasteners were reinstalled and the ductwork was secured.

However, licensee personnel did not complete a formal operability recommendation which would have required a more rigorous technical evaluation prior to declaring the fan operable. Instead the fan was declared operable and returned to service on June 20, based on various data obtained and visual observations. Consequently, the fan was returned to service with more significant motor bearing degradation than identified by licensee personnel and the bearing subsequently failed on July 1.

Therefore, the identification of Fan V-4As degraded condition was not complete and accurate, and this problem was not identified in a timely manner commensurate with its significance. The fact that the initial operability determination was non-conservative was recognized by licensee personnel and appropriately documented in the root cause evaluation.

This issue was considered a finding of very low safety significance (Green) and enforcement discretion was required to complete repairs to Fan V-4A with the plant at power which are discussed in detail in Sections 1R15 and 4OA5, respectively, of this report.

.4 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 condition reports were being generated and entered into the corrective action program with the appropriate significance characterization. For select condition reports, the inspectors also verified that identified corrective actions were reasonable, and had been implemented or were scheduled to be implemented in a manner commensurate with the significance of the identified problem. The condition reports that the inspectors reviewed are included in the list of documents for the specific inspection activities which is attached to this report.

b. Findings

No findings of significance were identified.

4OA3 Event Follow-up

.1 Grid Disturbance on August 14, 2003

a. Inspection Scope

On August 14, 2003, the inspectors observed plant parameters and equipment status during the automatic start of Emergency Diesel Generators 1-1 and 1-2 to verify that the plant equipment responded as designed to a grid disturbance event. The inspectors provided continuous 24-hour site coverage to monitor plant activities during the grid disturbance. Emergency Diesel Generators 1-1 and 1-2 automatic start was caused by the voltage drop on the 2140 Volt safety busses due to the loss of the electrical transmission grid on portions of the east coast of the United States and parts of Canada. Licensee operator response during this event was assessed under Personnel Performance Related to Non-Routine Plant Evolutions and Events, in Section 1R14.2 of this report.

b. Findings

No findings of significance were identified.

4OA4 Cross-Cutting Aspects of Findings

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

action deficiency, in that, the lack of rigor in an operability determination for Containment Air Cooler Fan V-4A failed to identify the extent of fan motor bearing degradation.

Consequently, Fan V-4A subsequently tripped on July 1, 2003, because of a failed motor bearing after the fan had been declared operable and returned to service on June 20, 2003. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified.

4OA5 Other Activities

(Closed) Unresolved Item (URI) 05000255/2003006-06: Review of Notice of Enforcement Discretion (NOED) 03-3-005 For Nuclear Management Company LLC Regarding Palisades The inspectors reviewed the circumstances associated with issuing NOED 03-3-005 and the basis for the NOED request to determine if a failure to comply with regulatory requirements contributed to the need for enforcement discretion. The inspectors also verified that licensee personnel complied with the compensatory actions contained in the NOED.

On July 1, 2003, at 4:14 a.m., Containment Air Cooler Fan V-4A tripped unexpectedly.

Technical Specification 3.6.6, Containment Cooling Systems, Condition A was entered which required that with one or more containment cooling trains inoperable, restore the train(s) to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Technical Specification 3.6.6, Condition B required that if Condition A could not be met, then be in Mode 3 (Hot Standby) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Mode 4 (Hot Shutdown) within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Licensee personnel determined that the repairs necessary to return Fan V-4A to an operable status would require more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Consequently, enforcement discretion would be needed to complete the repairs with the plant at power and preclude a plant shutdown to Mode 3. On July 3, 2003, licensee personnel requested enforcement discretion for the 72-hour completion time specified by TS 3.6.6, Condition A for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to complete the repairs to restore Containment Air Cooler Recirculation Fan V-4A to an operable status and preclude a plant shutdown.

The NRC verbally granted NOED 03-3-005 at 1:37 p.m. on July 3, 2003. Licensee personnel subsequently replaced the motor on Fan V-4A and declared Fan V-4A operable on July 6, 2003, at 2:04 p.m., which was within the completion time approved in the NOED.

No findings of significance were identified during the inspectors review of the basis of the NOED request and the licensee's implementation of compensatory actions required by the NOED. This URI is closed.

This issue was determined to be a self-revealed finding which is discussed further in Section 1R15 of this report.

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. D. Malone and other members of licensee management on October 9, 2003. Licensee personnel acknowledged the findings presented. The inspectors asked licensee personnel whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

.2 Interim Exit Meetings

The following Interim Exit Meeting was conducted:

C Occupational Radiation Safety ALARA and access control programs inspection with Mr. D. Cooper on August 29, 2003.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

D. Cooper, Senior Vice President of Group Operations

D. J. Malone, Site Vice President

M. Carlson, Engineering Director

P. Harden, Site Director

D. G. Malone, Supervisor, Regulatory Assurance

G. Packard, Operations Manager

R. Remus, Plant Manager

D. Williams, Manager - Chemistry and Radiation Protection

C. Moeller, ALARA Supervisor

Nuclear Regulatory Commission

D. Hood, Project Manager, NRR

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000255/2003006-01 NCV Degraded Motor Bearing in Containment Air Cooler Fan V-4A

05000255/2003006-02 URI Failure to Test the ATWS System

05000255/2003006-03 NCV Failure to Obtain a Radiological Briefing Prior to Entry Into a High Radiation Area

05000255/2003006-04 NCV Failure to Meet Radiation Work Permit Requirements Upon Receipt of an Electronic Dosimetry Alarm

05000255/2003006-05 NCV Failure to Barricade and Post a High Radiation Area (Section 2OS5.1)

05000255/2003006-06 URI Review of Notice of Enforcement Discretion 03-3-005 (Section 4OA5)

ATTACHMENT

Closed

05000255/2003006-01 NCV Degraded Motor Bearing in Containment Air Cooler Fan V-4A

05000255/2003006-03 NCV Failure to Obtain a Radiological Briefing Prior to Entry Into a High Radiation Area

05000255/2003006-04 NCV Failure to Meet Radiation Work Permit Requirements Upon Receipt of an Electronic Dosimetry Alarm

05000255/2003006-05 NCV Failure to Barricade and Post a High Radiation Area (Section 2OS5.1)

05000255/2003006-06 URI Review of Notice of Enforcement Discretion 03-3-005 (Section 4OA5)

Discussed

None ATTACHMENT