IR 05000255/2006013

IR 05000255-06-013; 10/01/2006 - 12/31/2006; Palisades Nuclear Plant; Surveillance Testing, as Low as Is Reasonably Achievable Planning and Controls, and Other Activities

ML070240635
Person / Time
Site:	Palisades
Issue date:	01/24/2007
From:	Christine Lipa NRC/RGN-III/DRP/RPB4
To:	Harden P Nuclear Management Co
References
50-255, 72-7 (2.206), FOIA/PA-2010-0209, RAS 13891 IR-06-013
Download: ML070240635 (44)

Text

ary 24, 2007

SUBJECT:

PALISADES NUCLEAR PLANT NRC INTEGRATED INSPECTION REPORT 05000255/2006013

Dear Mr. Harden:

On December 31, 2006, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Palisades Nuclear Plant. The enclosed report documents the inspection findings which were discussed on January 10, 2007, 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, one NRC-identified finding and two self-revealed findings of very low safety significance (Green) were identified. Two of these findings were determined to involve violations of NRC requirements. Additionally, two licensee-identified violations which were determined to be of very low safety significance are described in this report. However, because the violations were of very low safety significance and because the issues have been entered into your corrective action program, the NRC is treating these findings as a non-cited violations (NCVs) consistent with Section VI.A.1 of the Enforcement Policy.

If you contest the subject or severity of an NCV, you should provide a response with a basis for your denial, within 30 days of the date of this inspection report, 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, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Palisades facility. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Christine A. Lipa, Chief Branch 4 Division of Reactor Projects Docket Nos. 50-255 and 07200007 License No. DPR-20 Enclosure: Inspection Report 05000255/2006013 w/Attachment: Supplemental Information cc w/encl: M. Sellman, President and Chief Executive Officer R. Fenech, Senior Vice President, Nuclear Fossil and Hydro Operations D. Cooper, Senior Vice President - Group Operations L. Lahti, Manager, Regulatory Affairs J. Rogoff, Vice President, Counsel and Secretary A. Udrys, Esquire, Consumers Energy Company S. Wawro, Director of Nuclear Assets, Consumers Energy Company Supervisor, Covert Township Office of the Governor State Liaison Office, State of Michigan L. Brandon, Michigan Department of Environmental Quality -

Waste and Hazardous Materials Division

SUMMARY OF FINDINGS

IR 05000255/2006013; 10/01/2006 - 12/31/2006; Palisades Nuclear Plant; Surveillance Testing,

As Low As Is Reasonably Achievable (ALARA) Planning and Controls, and Other Activities.

This report covers a 3-month period of baseline inspections. The inspections were conducted by Region III inspectors and resident inspectors. This report includes three Green findings, two of which were non-cited violations (NCVs). 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 Nuclear Regulatory Commission (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.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

A Green self-revealing NCV of 10 CFR 50, Appendix B, Criterion VIII,

"Identification and Control of Materials, Parts and Components" was identified for failing to have adequate control measures needed to prevent the use of defective parts.

Specifically, a fuel leak developed due to the incorrect part on the 1-2 Emergency Diesel Generator (EDG) on November 20, 2005, that resulted in aborting a surveillance test.

The cause was related to a defective part which had been installed 28 days earlier. The part has been replaced, and there are no other susceptible parts in the diesel engines on site.

The finding is more than minor since the defective part impacted the cornerstone for availability, reliability and capability of the class 1E, on site EDG system and is an associated attribute of equipment performance. The finding screened as very low safety significance, Green, since there was no loss of safety function for the 1-2 EDG.

(Section 4OA5)

Cornerstone: Barrier Integrity

Green.

The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion XI, "Test Control" for the failure to have an accurate Technical Specification (TS) surveillance procedure for primary coolant leakage measurement. Specifically, the licensee did not provide an accurate calculation or accurate acceptance criteria over all the temperature ranges and other plant conditions under which the surveillance procedure could be used. This issue was entered into the licensees corrective action system and the licensee developed interim guidance on leak rate calculations pending a procedure revision.

The finding is more than minor because it can reasonably be viewed as a precursor to a more significant event because the errors can prevent recognition of leakage in excess of the TS and licensing basis. The finding screened as very low safety significance,

Green, using the Phase 1 worksheet of IMC 0609, Appendix A, since no actual cases were found where unidentified leakage exceeded the TS. (Section 1R22)

Cornerstone: Occupational Radiation Safety

Green.

A Green finding was self-revealed for failure to adequately implement radiological dose controls during Refueling Outage 18 (RO18). Specifically, work control and planning issues (worker fatigue, worker proficiency, and material condition)contributed to additional worker doses. The total sum of the occupational radiation doses (collective dose) received by individuals for one work activity was found in excess of that collective dose planned or intended (i.e., that dose the licensee determined was ALARA for those work activities).

The finding was more than minor because the issue was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of the worker<s health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors concluded that the finding did not result in an occupational overexposure, a substantial potential for an overexposure, or a compromised ability to assess dose. The inspectors determined that the finding involved ALARA planning and work controls. Considering the licensees current 3-year rolling collective dose average exceeds 135 person-rem per unit, the actual dose was less than 25 person-rem and there are no other occurrences, the inspectors concluded that the SDP assessment for this finding was of very low safety significance,

Green.

The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because the licensee failed to appropriately coordinate work activities. (Section 2OS2)

Licensee-Identified Violations

Two violations of very low safety significance, which were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. These violations and corrective actions are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

The plant began the inspection period at or near full rated thermal power and operated at full power until November 1, 2006. On November 1, 2006, a through-wall leak was discovered on a containment air cooler and the licensee shutdown to mode 3 as required by Technical Specifications (TS). The licensee repaired the air cooler with a temporary modification and restarted on November 3, 2006. On November 5, 2006, the plant returned to near full rated thermal power and remained at or near full rated thermal power for the rest of the inspection period.

REACTOR SAFETY

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

1R01 Adverse Weather Protection

.1 Adverse Weather Preparation

a. Inspection Scope

The inspectors verified that licensee personnel implemented the appropriate actions for very high winds at the site on October 13, 2006. This included review of the plants response and entry into Off-Normal Procedure 12, "Acts of Nature." The inspectors interviewed personnel, walked-down affected areas and reviewed the licensing and design basis for structures and equipment. This is considered one site sample.

b. Findings

No findings of significance were identified.

.2 Preparation for Cold Weather

a. Inspection Scope

The inspectors reviewed the plants preparation for cold weather. The inspectors used the Updated Final Safety Analysis Report (UFSAR), TS, plant procedures, and past adverse conditions and corrective actions to assess systems that could be adversely affected by cold weather. The inspectors performed a walkdown of susceptible systems. The inspectors also reviewed the licensees cold weather procedures. The safety system focus was on the safety injection system from the Safety Injection and Refueling Water Tank (SIRWT); risk significant systems, structures, and components (SSCs) including the EDG 1-3; and the Auxiliary Feedwater (AFW) system. The documents reviewed during this inspection are listed in the attachment. This constitutes one system sample.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

Partial Walkdowns (71111.04Q)

a. Inspection Scope

The inspectors completed three equipment alignment inspection samples by performing partial walkdowns on the following risk-significant plant equipment:

• Right train High Pressure Safety Injection (HPSI) equipment during a planned outage on HPSI pump P 66B (October 11, 2006)

• EDG 1-2 with EDG 1-1 out of service for planned maintenance (October 31, 2006)

• Motor driven AFW pumps during a turbine driven AFW pump planned outage (November 15, 2006)

During the walkdowns, the inspectors verified that power was available, accessible equipment and components were appropriately aligned, and no open work orders for known equipment deficiencies existed which would impact system availability.

The inspectors also reviewed selected condition reports related to equipment alignment 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 and implemented as scheduled. The documents reviewed during this inspection are listed in the attachment.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Fire Area Walkdowns

a. Inspection Scope

The inspectors completed seven fire protection inspection samples by touring the following areas in which a fire could affect safety-related equipment:

C Turbine Building 590 foot elevation (Fire Area 23D)

C Battery Room for ED01 battery (Fire Area 12)

C Battery Room for ED02 battery (Fire Area 11)

C Turbine Building 611 foot elevation (Fire Are 23 D)

C Component Cooling Water (CCW) Roof and SIRWT (Fire Area 32)

C EDG 1-1 Room (Zone 5)

C EDG 1-2 Room (Zone 6)

The inspectors verified that transient combustibles and ignition sources were appropriately controlled, and that the installed fire protection equipment in the fire areas corresponded with the equipment that was referenced in the UFSAR, Section 9.6, "Fire Protection." The inspectors also assessed the material condition of fire suppression systems, manual fire fighting equipment, smoke detection systems, fire barriers and emergency lighting units. For selected areas, the inspectors reviewed documentation for completed surveillances to verify that fire protection equipment and fire barriers were tested as required to ensure availability.

The inspectors reviewed selected condition reports associated with fire protection to verify that identified problems were entered into the corrective action program with the appropriate significance characterization. The inspectors also verified that planned and completed corrective actions were appropriate. The documents reviewed during this inspection are listed in the attachment.

b. Findings

No findings of significance were identified.

.2 Fire Protection - Drill Observation and Annual Inspection

a. Inspection Scope

The inspectors observed an unannounced fire drill on October 24, 2006, to evaluate the fire brigades performance. The inspectors observed the brigades response to the fire to verify timeliness, proper donning of equipment, and command and control by the brigade leader. In addition, the inspectors reviewed procedures, fire fighting equipment, breathing air requirements, and corrective action for adverse conditions. Finally the inspectors, on a sampling basis, verified that members assigned to the fire brigade met the qualification requirements. The inspectors evaluated the licensees critique of the drill and actions taken as a result of the critique. Specific attributes evaluated are listed in Fire Protection, IP 71111.05AQ, paragraph 02.02. This constituted one sample.

b. Findings

No findings of significance were identified.

1R06 Flood Protection

a. Inspection Scope

The inspectors completed one inspection sample pertaining to flood protection measures for internal flooding events. The inspectors performed a walkdown of the intake structure and the EDG rooms and associated flood barriers to verify the flood barriers were in acceptable condition. The intake structure contains all three safety-related service water (SW) pumps for response to various events. The inspectors reviewed the licensees flood analysis and licensing basis for the areas to determine if the analysis was consistent with configuration of the room. Further, the inspectors reviewed condition reports to verify that corrective actions for previously identified flood protection problems were appropriate and had been properly implemented.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

a. Inspection Scope

The inspectors completed one inspection sample of licensed operator requalification training by observing a crew of licensed operators during simulator training on October 25, 2006. The inspectors assessed the operators response to the simulated events which included a loss of primary coolant.

The inspectors verified that the operators were able to effectively mitigate the events through accurate and timely implementation of applicable alarm response procedures; Off-Normal Procedures and Emergency Operating Procedures. The inspectors also observed the post-training critique to assess the licensee evaluators and the crews ability to self-identify performance deficiencies.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors completed four inspection samples pertaining to maintenance effectiveness by reviewing maintenance rule implementation activities for the following systems and components:

• Control Air and Safety-Related Instrument Air

• Critical SW

• Containment Air Cooling

• Shutdown Cooling and Low Pressure Safety Injection The inspectors reviewed the licensee's implementation of the maintenance rule requirements to verify that component and equipment failures were evaluated and appropriately dispositioned. The inspectors also verified that the selected systems and components were scoped into the maintenance rule and properly categorized as (a)(1) or (a)(2) in accordance with 10 CFR 50.65.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

The inspectors completed five inspection samples. The inspectors reviewed the following activities to verify that the appropriate risk assessments were performed prior to removing equipment for work. The inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors verified the appropriate use of the licensees risk assessment tool and risk categories in accordance with Administrative Procedure 4.02, Control of Equipment, Revision 36, and Fleet Procedure FP-OP-RSK-01, Risk Monitoring and Risk Management, Revision 0. Documents reviewed are listed in the attachment.

• Elevated risk during a planned HPSI outage during the week of October 8, 2006

• Elevated risk due to planned work on a condensate fill valve

• Elevated risk due to a 4 day planned outage on 1-1 EDG

• Elevated risk due to SW pump P-7B repack

• Risk due to severe weather and scheduled EDG testing on December 1, 2006 The inspectors also verified that condition reports related to emergent equipment problems were entered into the corrective action program with the appropriate significance characterization. Select condition reports related to risk management during maintenance activities were reviewed to verify that planned corrective actions were appropriate and had been implemented as scheduled.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors completed six inspection samples. For the six Operability Recommendations (OPRs) listed below, the inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors reviewed the UFSAR to verify that the system or component remained available to perform its intended function. In addition, the inspectors reviewed compensatory measures implemented to verify that the compensatory measures worked as stated and the measures were adequately controlled. In addition, the inspectors verified that the condition reports generated for equipment operability issues were entered into the licensees corrective action program with the appropriate significance characterization. Documents reviewed are listed in the attachment.

• Operability recommendation on the EDGs due to autostart of fans

• Corrosion and loss of terminal plating material on Cell 25 of Battery ED02

• HPSI OPR on nonconformance with casing alignment blocks

• Control Circuit Voltage Drop Discrepancy for 480V Motor Control Centers

• VHX-4 Containment Air Cooler Had Three Through-wall Leaks in the Last 13 Months

• CV0823, High Capacity SW Water Flow to CCHX, Valve Position Indication Not Functioning

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications

a. Inspection Scope

The inspectors completed one inspection sample of a permanent plant modification package that involved isolating instrument air to two SW supply valves. The inspectors reviewed the design change information, related design basis documents and the 10 CFR 50.59 screening evaluation to verify that the design bases, licensing bases and performance capability of the involved diesel generator system were not degraded by this modification. In addition, the inspectors reviewed applicable plant documents to verify that any appropriate changes were made. Documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

The inspectors completed four inspection samples pertaining to post maintenance testing by assessing testing activities that were conducted for the following maintenance activities:

• Testing of CCW pump P-52A following maintenance

• Testing of EDG 1-2 after troubleshooting for low fuel oil pressure

• Testing of VHX-4 following leak repair

• Testing of the P-7C SW pump following inspection for debris intrusion The inspectors observed portions of the post maintenance testing and/or reviewed documentation to verify that the tests were performed as prescribed by the work orders and test procedures; 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 reviewed documentation to verify that the test criteria and acceptance criteria were appropriate for the scope of work performed; reviewed test procedures to verify that the tests adequately verified system operability; and reviewed documented test data to verify that the data was complete, and that the equipment met the prescribed acceptance criteria. Further, the inspectors reviewed condition reports to verify that post maintenance testing problems were entered into the corrective action program with the appropriate significance characterization. For select condition reports, the inspectors verified that the corrective actions were appropriate and implemented as scheduled.

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

Unplanned Forced Outage

a. Inspection Scope

The inspectors observed and assessed the licensees performance in completing activities during an unplanned outage that occurred due to a through-wall leak on VHX-4 SW piping. The outage lasted from November 1 to November 3, 2006. This inspection constitutes one sample. The inspectors performed the following activities periodically throughout the outage:

• Verified that plant equipment, including inventory control systems, required to minimize plant risk were aligned in accordance with plant procedures

• Reviewed selected condition reports to verify that identified problems were accurately characterized; entered into the corrective action program with the appropriate significance; and that corrective actions were appropriate

• Verified decay heat removal systems were aligned per TS

• Randomly verified that station electrical power, emergency diesel generators, decay heat removal and primary coolant system inventory control systems were aligned as required

• Observed operator performance during portions of reactor shutdown, startup, and power ascension

b. Findings

During reactor startup, the inspectors noted that all AFW pumps were in the manual mode of operation and would not have started automatically on a steam generator low level as required by TS. The licensee corrected this issue and initiated an event response and root cause investigation. The NRC determined a Special Inspection Team (SIT) was required to investigate the issue and NRC Inspection Report 05000255/2006014 documents the findings of that team.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors witnessed four surveillance tests and/or reviewed test data of selected risk-significant SSCs, listed below, to assess, as appropriate, whether the SSCs met the requirements of the TS; the UFSAR; Palisades Administrative Procedure 9.20; TS Surveillance and Special Testing Program; Engineering Manual EM-09-02 and EM-09-04, Inservice Testing of Plant Valves and Inservice Testing of Selected Safety-Related Pumps. One of the samples was an inservice test and one sample was the Primary Coolant System (PCS) leakrate procedure. The inspectors also determined whether the testing effectively demonstrated that the SSCs were operationally ready and capable of performing their intended safety functions. Further, the inspectors reviewed selected condition reports regarding surveillance testing activities. The inspectors verified that the identified problems were entered into the licensees corrective action program with the appropriate significance characterization and that the planned and completed corrective actions were appropriate. Additional documents reviewed are listed in the attachment.

C Inservice testing of SW pump P-7B C PCS Leakrate Surveillance DWO-1, October 11, 2006 C Diesel Fuel Oil Testing for Ultra Low Sulfur Diesel Fuel C Terminal resistance surveillance for safety-related batteries, October 18, 2006

b. Findings

Introduction:

The inspectors identified a finding of very low safety significance associated with a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" for failing to have an accurate TS surveillance procedure to determine PCS water inventory balance. Specifically, the licensee did not provide an accurate calculation or accurate acceptance criteria over all the temperature ranges and other plant conditions under which the surveillance procedure was applicable.

Description:

The inspectors reviewed the PCS leakage calculation contained in DWO-1, Technical Surveillance Procedure Operator Daily/Weekly Items Modes 1, 2, 3, and 4.

As part of the review, the inspectors evaluated corrective actions related to a previously NRC identified deficiency captured in CAP 047198 on March 22, 2005. The Corrective Action Program (CAP) documented the lack of an engineering analysis for the calculation in the procedure. Although over a year old, corrective actions remained open.

During the inspectors review of the calculation, the inspector noted that the calculation used a single constant to correct for PCS temperature of 74.43 gallons per degree Fahrenheit (F). Since water density varies with temperature, and pressure to a lessor degree, the inspectors calculated potential errors associated with this oversight. Over the temperature and pressure ranges the calculation applies, from Modes 1 through 4, the inspectors calculated leakrate errors as high as

.7 gallons per minute (gpm). In

Mode 1, the error would be limited to

.02 gpm and a .15 gpm error could exist in Mode

3. The surveillance verifies compliance with TS leakage limits of 1 gpm as well as a determinant for entry into off-normal procedures for a primary coolant leak. In addition, the inspectors noted that the calculation had not been adjusted to account for plugging of steam generator (SG) u-tubes. This issue was entered into the licensees corrective action system and the licensee developed interim guidance on leak rate calculations pending a procedure revision.

Analysis:

The inspectors concluded that not having an accurate calculation to determine compliance with TS surveillance criteria was a performance deficiency which required a significance evaluation. The issue is more than minor because the finding can reasonably be viewed as a precursor to a more significant event. The leak rate determination satisfies TS surveillance for monitoring of PCS leakage. In addition, leak rate monitoring provides both the licensee and the NRC with warning of incipient challenges to PCS integrity. The magnitude of errors associated with the calculation can reasonably mask incipient failures as well as leaks in excess of TS requirements.

The issue screens as very low safety significance since there were no cases of leakrates approaching the TS limit in the last year based on documents reviewed; and no cases where the inaccurate calculation impacted the plants response to unidentified leakage. Therefore, this finding screened as Green using the Phase 1 worksheet of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations."

Enforcement:

10 CFR Part 50, Appendix B, Criterion XI, "Test Control" requires, in part, that tests on structures, systems and components are performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The PCS is a safety-related system which has acceptance limits contained in TS 3.4.13b and is tested in accordance with SR 3.4.13.1. DWO-1 is the licensee procedure which accomplishes this test. Contrary to the above, DWO-1, Technical Surveillance Procedure Operator Daily/Weekly Items Modes 1, 2, 3, and 4, Revision 73 did not have adequate acceptance limits which addressed the accuracy of the calculation for possible conditions for which this test could be used. However, because this violation was of very low safety significance and because the issue was entered into the licensee's corrective action program (AR01055966) this violation is being treated as an NCV, consistent with Section VI.A.1 of the Enforcement Policy (NCV 05000255/2006013-01). The licensee implemented operational guidance to bound use of the existing calculation until a procedure revision could be completed.

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors completed one inspection sample by reviewing the following temporary modification:

• EC 8866, "Temporary Modification to T-2 Condensate Storage Tank Lid" The condensate storage tank was accidently subjected to an over pressure event during a plant transient many years ago that resulted in a local failure of the wall-to-roof seal weld. The failed welded joint occurred between the walls angled member and roof sheet metal causing a local distortion and rupture about 3 inches wide and 57 inches long at the 3/16 inch wall-to-roof fillet weld. This temporary modification was installed to partially improve the structural capability in the failed weld region of the tank from sustained 100 miles per hour winds as described in Section 5.3 of the UFSAR for Class 1 structures. The inspectors interviewed engineering department personnel and reviewed the design documents and applicable 10 CFR 50.59 evaluation to verify that TS and the UFSAR requirements were satisfied. The inspectors also conducted a walkdown of the installation to verify that the modification was implemented as designed and that the modification did not adversely impact auxiliary feedwater system operability or availability.

b. Findings

No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspectors completed screening reviews of Revisions 14 and 15 of the Palisades Nuclear Plants Emergency Plan to determine whether changes identified in these revisions may have reduced the effectiveness of the licensees emergency planning.

The screening reviews of these revisions do not constitute approval of the changes and, as such, the changes are subject to future NRC inspection to ensure that the Emergency Plan continues to meet NRC regulations.

These activities completed one inspection sample.

b. Findings

No findings of significance were identified.

1EP6 Emergency Preparedness Drill Evaluation

a. Inspection Scope

Resident inspectors evaluated the conduct of two routine licensee simulator scenarios on November 1, 2006 and November 15, 2006, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. This scenario was part of the licensees planned Drill/Exercise Performance evaluation. The inspectors observed emergency response operations in the simulated control room to verify that event classification and notifications were properly completed in accordance with E1-1 Emergency Plan Classification Matrix, and Site Emergency Plan (SEP), Revision 15. The inspectors also attended the licensee critique of the drill to compare any inspector-observed weakness with those identified by the licensee in order to verify whether the licensee was properly identifying failures.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS2 ALARA Planning and Controls (71121.02)

.1 Radiological Work Planning

a. Inspection Scope

The inspectors compared the results achieved (dose rate reductions, person-rem used)with the intended dose established in the licensees ALARA planning for Refueling Outage 18 (RO18) work activities. The inspectors reviewed the reasons for any inconsistencies between intended and actual work activity doses. This review represents one sample.

The inspectors compared the person-hour estimates provided by maintenance planning and other groups to the radiation protection group with the actual work activity time requirements and evaluated the accuracy of these time estimates. This review represents one sample.

The inspectors assessed whether post-job (work activity) reviews were conducted and if identified problems were entered into the licensees corrective action program. This review represents one sample.

b. Findings

No findings of significance were identified.

.2 Verification of Dose Estimates and Exposure Tracking Systems

a. Inspection Scope

The inspectors reviewed the assumptions and basis for the current annual collective exposure estimate. The inspectors reviewed applicable procedures to determine the methodology for estimating work activity-specific exposures and the intended dose outcome. The inspectors evaluated both dose rate and man-hour estimates for reasonable accuracy. This review represents one sample.

The inspectors reviewed the licensees method for adjusting exposure estimates, or re-planning work, when unexpected changes in scope or emergent work are encountered. This review represents one sample.

The inspectors reviewed the licensees exposure tracking system. The inspectors assessed whether the level of exposure tracking detail, exposure report timeliness and exposure report distribution is sufficient to support control of collective exposures. This review represents one sample.

b. Findings

One finding of very low safety significance was identified as described below:

Introduction:

One self-revealing Green finding was identified associated with the failure to adequately implement radiological dose controls during RO18. Specifically, work control and planning issues (worker fatigue, worker proficiency, and material condition)contributed to additional worker doses. The collective dose for one work activity (26400; Reactor Head Insulation) resulted in actual dose in excess of 5 person-rem and also exceeding the initial planned dose estimates by more than 50 percent.

Description:

The initial dose estimates for Work Order 26400 was primarily based on historical dose rates for the same or similar work activity and person-hour estimates provided by the maintenance groups responsible for the evolution. Additional requirements were established to conduct mock-up training to develop worker proficiency to meet historic performance times. However, the mock-up training did not include the removal of installed lead shielding, an activity that was performed by another work group during previous outages. This work activity fatigued the insulation crew and impacted the ability to perform insulation activities similar to historic values. The initial Work Order dose estimate of 4.107 rem was reviewed by station management. The activity was completed for 6.451 rem. The ALARA in-progress, post-job reviews, condition reports, and personal interviews conducted at the station identified three work control and/or planning issues to explain the differences between the initially projected and actual doses received. The causes were:

• Worker fatigue

• Materiel condition of the insulation

• Worker proficiency Neither the inspectors nor the licensee identified any significant changes to dose rates in the area or to the originally planned work scope to this Work Order.

Analysis:

The failure to adequately implement radiological dose controls represents a performance deficiency as defined in IMC 0612, "Power Reactor Inspection Reports,"

Appendix B, "Issue Screening." The inspectors determined that the issue was associated with the Program & Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Therefore, the issue was more than minor and represented a finding which was evaluated using the SDP.

Since this finding involved radiological controls and ALARA planning, the inspectors utilized IMC 0609, Appendix C, "Occupational Radiation Safety SDP," to assess its significance. The inspectors concluded that the finding did not result in an occupational overexposure, a substantial potential for an overexposure, or a compromised ability to assess dose. The inspectors determined that the finding involved ALARA Planning and work controls. Considering that the licensees current 3-year rolling collective dose average exceeds 135 person-rem per unit, the actual dose was less than 25 person-rem and there were no other occurrences, the inspectors concluded that the SDP assessment for this finding was of very low safety significance (Green). The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because the licensee failed to appropriately coordinate work activities.

Enforcement:

The failure to adequately implement radiological dose control was a performance deficiency under the reactor oversight process (ROP); however, no violation of regulatory requirements occurred. This issue (FIN 05000255/2006013-02)was considered a finding of very low safety significance and is documented in the licensees corrective action program (AR 01023058, AR 01042960, and AR 01045001).

Corrective actions include adding outage planners that could actively follow fewer activities.

.3 Job Site Inspections and ALARA Control

a. Inspection Scope

The inspectors reviewed exposures of individuals from selected work groups. The inspectors evaluated any significant exposure variations which may exist among workers and assessed whether these significant exposure variations are the result of worker job skill differences or whether certain workers received higher doses because of poor ALARA work practices. This review represents one sample.

b. Findings

No findings of significance were identified.

.4 Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed the licensees self-assessments, audits, and Special Reports related to the ALARA program since the last inspection. The inspectors assessed whether the licensees overall audit programs scope and frequency (for all applicable areas under the Occupational Cornerstone) met the requirements of 10 CFR 20.1101(c).

The inspectors assessed whether identified problems were entered into the corrective action program for resolution. The inspectors reviewed dose significant post-job (work activity) reviews and post-outage ALARA report critiques of exposure performance. The inspectors assessed whether identified problems were properly characterized, prioritized, and resolved in an expeditious manner.

The inspectors reviewed corrective action reports related to the ALARA program. The inspectors interviewed staff and reviewed documents to assess whether the follow-up activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk:

• 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 NCVs tracked in corrective action system

• Implementation/consideration of risk significant operational experience feedback The inspectors review placed emphasis on ensuring problems were identified, characterized, prioritized, entered into a corrective action, and resolved. For repetitive deficiencies or significant individual deficiencies in problem identification and resolution identified above, the inspectors evaluated if the licensees self-assessment activities were also identifying and addressing these deficiencies. These reviews represented one sample.

b. Findings

No findings of significance were identified.

2PS2 Radioactive Material Processing and Transportation (71122.02)

.1 Radioactive Waste System Inspection Planning

a. Inspection Scope

The inspectors reviewed the liquid and solid radioactive waste system description in the UFSAR for information on the types and amounts of radioactive waste (radwaste)generated and disposed. The inspectors reviewed the scope of the licensees audit program with regard to radioactive material processing and transportation programs to verify that it met the requirements of 10 CFR 20.1101(c). This review represented one sample.

b. Findings

No findings of significance were identified.

.2 Walkdown of Radioactive Waste Systems

a. Inspection Scope

The inspectors reviewed the liquid and solid radioactive waste system description in the UFSAR and the most recent information regarding the types and amounts of radioactive waste generated and disposed. The inspectors performed walkdowns of the liquid and solid radwaste processing systems to verify that the systems agreed with the descriptions in the UFSAR and the Process Control Program and to assess the material condition and operability of the systems. The inspectors reviewed changes to the waste processing system to verify the changes were reviewed and documented in accordance with 10 CFR 50.59 and to assess the impact of the changes on radiation dose to members of the public.

The inspectors reviewed the current processes for transferring waste resins into transportation containers to determine if appropriate waste stream mixing and/or sampling procedures were utilized. The inspectors also reviewed the methodologies for waste concentration averaging to determine if representative samples of the waste product were provided for the purposes of waste classification in accordance with 10 CFR 61.55. During this inspection, the licensee was not conducting waste processing. This review represented one sample.

b. Findings

No findings of significance were identified.

.3 Waste Characterization and Classification

a. Inspection Scope

The inspectors reviewed the licensees radiochemical sample analysis results for each of the licensees waste streams, including dry active waste, resins, and filters. The inspectors also reviewed the licensees use of scaling factors to quantify difficult-to-measure radionuclides (e.g., pure alpha or beta emitting radionuclides). The reviews were conducted to verify that the licensees program assured compliance with 10 CFR 61.55 and 10 CFR 61.56, as required by Appendix G of 10 CFR Part 20. The inspectors also reviewed the licensees waste characterization and classification program to ensure that the waste stream composition data accounted for changing operational parameters and thus remained valid between the annual sample analysis updates. This review represented one sample.

b. Findings

No findings of significance were identified.

.4 Shipment Preparation

a. Inspection Scope

The inspectors reviewed shipment packaging, surveying, labeling, marking, placarding, vehicle checks, emergency instructions, disposal manifest, shipping papers provided to the driver, and licensee verification of shipment readiness for a dry active waste shipment. The inspectors verified that the receiving licensee was authorized to receive the shipment packages. The inspectors reviewed the licensees procedures for loading and closure. The inspectors observed radiation worker practices to verify that the workers had adequate skills to accomplish each task and to determine if the shippers were knowledgeable of the shipping regulations and whether shipping personnel demonstrated adequate skills to accomplish the package preparation requirements for public transport with respect to NRC Bulletin 79-19 and 49 CFR Part 172 Subpart H.

The inspectors reviewed the training provided to personnel responsible for the conduct of radioactive waste processing and radioactive shipment preparation activities. The review was conducted to verify that the licensees training program provided training consistent with NRC and Department of Transportation requirements. This review represented one sample.

b. Findings

No findings of significance were identified.

.5 Shipping Records

a. Inspection Scope

The inspectors reviewed ten non-excepted package shipment manifests completed in years 2005 and 2006 to verify compliance with the NRC and Department of Transportation requirements (i.e., 10 CFR Parts 20 and 71 and 49 CFR Parts 172 and 173). Since no ongoing shipping activities were occurring, the inspector was not able to review current package preparations or shipping during the inspection. This review represented one sample.

b. Findings

No findings of significance were identified.

.6 Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed audits and self-assessments that addressed radioactive waste and radioactive materials shipping program deficiencies since the last inspection, to verify that the licensee had effectively implemented the corrective action program and that problems were identified, characterized, prioritized and corrected. The inspectors also verified that the licensee's self-assessment program was capable of identifying repetitive deficiencies or significant individual deficiencies in problem identification and resolution.

The inspectors also reviewed corrective action reports from the radioactive material and shipping programs since the previous inspection, interviewed staff and reviewed documents to determine if the following activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk:

• 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 NCVs tracked in corrective action system

• Implementation/consideration of risk significant operational experience feedback This review represented one sample.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

The inspectors sampled licensee submittals for the PIs listed below for the periods indicated. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in Revision 4 of Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," were used. The following PIs were reviewed:

Cornerstone: Initiating Events

The inspectors verified the Unplanned Transients per 7000 Critical Hours PI. The inspectors reviewed power history data from October 1, 2004, through September 30, 2006, determined the number of power changes greater than 20 percent full power that occurred, evaluated each of those power changes against the PI definition, and verified the licensee's calculation of critical hours. This constituted one inspection sample.

Cornerstone: Mitigating Systems

The inspectors verified the Safety System Functional Failures (SSFF) PI. The inspectors reviewed each Licensee Event Report (LER) from January 1, 2004, through September 30, 2006, determined the number of SSFF that occurred, evaluated each LER against the PI definitions, and verified the number of SSFF reported. This constituted one inspection sample.

Cornerstone: Barrier Integrity

Reactor Coolant System Leakage The inspectors reviewed leak rate data from operating logs and verified data submitted by the licensee and confirmed the licensee submitted accurate data. The inspectors looked at the submittals from 2005 through 2006 to verify the accuracy of the PI data.

The inspectors also reviewed the surveillance used to determine leak rate as documented in Section 1R22 of this report.

Reactor Coolant System Specific Activity The inspectors reviewed Chemistry Department records including isotopic analyses for 2005 through September 2006 to determine if the greatest dose equivalent iodine (DEI)values determined during steady state operations corresponded to the values reported to the NRC. The inspectors also reviewed selected DEI calculations including the application of dose conversion factors as specified in plant TS. Additionally, the inspectors accompanied a chemistry technician and observed the collection and preparation of reactor coolant system samples to evaluate compliance with the licensees sampling procedure protocols. Further, sample analyses and calculation methods were discussed with chemistry staff to determine their adequacy relative to TS, licensee procedures and industry guidelines.

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 condition reports were being generated and entered into the corrective action program (CAP) with the appropriate significance characterization. For select condition reports, the inspectors also verified that identified corrective actions were appropriate and had been implemented or were scheduled to be implemented in a timely manner commensurate with the significance of the identified problem.

b. Findings

No findings of significance were identified.

.2 Semi-Annual Trend Review to Identify Trends

a. Inspection Scope

As required by IP 71152, "Identification and Resolution of Problems," the inspectors performed a review of the licensees CAP action requests to identify trends that could indicate the existence of a more significant safety issue. The inspectors also reviewed the Operations Department Monthly Performance Report dated October 2006, the Site DRUM Report for the third Quarter of 2006 and the Corrective Action Program Performance Indicator Summary, dated November 2006. The inspectors review for potential trends included the results from the daily inspector CAP item screening discussed in Section 4OA2.1. The plant CAP action request screening meetings were observed to review the licensees level of effort in identifying potential trends, and any associated corrective actions that were planned or implemented. In addition, the inspectors reviewed issues documented outside the normal CAP that included, maintenance work orders, component status reports, performance indicators and Operations control room logs. The inspectors review nominally considered the 6 month period of July through December 2006. The inspectors compared and contrasted their results with the results obtained by the licensee during previous internal reviews. This constitutes one sample of trend reviews.

b. Findings

No findings of significance were identified.

.3 Annual Sample: Review of Barrier Controls for High Energy Line Breaks (HELB)

a. Inspection Scope

The inspectors reviewed the sites barrier control processes as they related to a postulated HELB. During baseline inspection activities, the inspectors questioned licensee HELB controls. In particular the inspectors noted the site did not have any HELB barriers designed to isolate turbine building HELBs from safety-related equipment. The inspectors reviewed the sites processes for control of HELBs and the licensing basis for HELB barrier control. In addition the inspectors reviewed Action Requests (ARs) related to HELB control including review of external operating experience. This constitutes one sample.

b. Unresolved Item (URI)

The inspectors identified an issue associated with two AFW pumps (8A and 8B) which are located in the turbine building. Specifically, the AFW pumps 8A and 8B are located in a room which is not protected from a harsh environment and the components are not rated for a harsh environment. Since the turbine building contains numerous high energy lines, an analysis of environmental conditions in the AFW room is required to determine the impact of a HELB in the turbine building on the AFW pumps in terms of temperature and humidity. The item has been placed into the corrective action process as CAP01068459. The third pump, 8C, is protected from a harsh environment and therefore there is no current safety issue. This is an URI pending completion of an assessment by the licensee and review by the NRC.

Description During a review of work activities associated with an AFW pump, the inspectors noted the watertight door to the AFW room was allowed to remain open during maintenance while the equipment in the room remained operable. The AFW room houses two of the three AFW pumps and resides in the turbine building. The third pump, 8C, is housed in the Auxiliary Building. The inspectors also noted that the exhaust duct for the room ventilation system is open to the turbine building interior with no HELB barrier. Finally, the non-safety ventilation system has damper alignments which can take a suction on the turbine building interior and discharge to the AFW room. The licensee noted the AFW room was not considered susceptible to a harsh environment. The original licensing basis document from Bechtel noted there was a large volume turbine building and therefore it is not likely to impact the AFW pump room. The report also noted at

.5 psi some building portions would give way to limit pressure. The inspectors, using

some basic calculations of heat transfer, demonstrated that, assuming an entire SG emptied into the turbine building, temperatures could exceed 200F. The inspectors calculations were rudimentary and did not consider Main Steam Line isolation or single failures; however, there are no other quantitative assessments of room temperature or humidity. The inspectors reviewed NRC Information Notice (IN) 2000-20, "Potential Loss of Redundant Safety-related Equipment Because of a Lack of High Energy Line Break Barriers." This IN is related to the impact on safety-related equipment due to failures of non safety-related high energy line piping. Although there are no barriers for the ventilation system for the 8A and 8B pumps, the potential impact is limited to those two pumps since the third pump is protected in the auxiliary building. Therefore, there is no current safety concern. Although there is no existing common failure mode, there is a reasonable basis to determine that a performance deficiency may exist; namely, that two AFW pumps are not protected from a harsh environment. This item will remain open as URI 05000255/2006013-03 until an assessment by the licensee is completed and reviewed by the NRC.

.4 Annual Sample: Review of Crane Issues

a. Inspection Scope

The inspectors reviewed the licensees causal analysis and corrective actions related to crane activities to evaluate how the licensee addressed NCVs05000255/2005012-01 and 05000255/2006004-04. The inspectors also reviewed other licensee CAPs related to cranes. Based on this review, the inspectors concluded that the licensee performed an adequate root cause and identified appropriate corrective actions for NCV 05000255/2005012-01 concerning improper manipulation of the fuel handling crane with a suspended load. However, the common cause for NCV 05000255/2006004-04 concerning the polar crane contacting and damaging a boom crane narrowly focused on prevention of the L-6 crane from contacting the L-1 crane. Despite an earlier event where the crane operator contacted a SG with a load, the licensee did not consider broad causal factors, operating experience, or corrective actions such that corrective actions would broadly address human errors in crane operation. However, the licensee has site wide performance improvement initiatives in progress that in part address crane operations.

b. Findings

No findings of significance were identified.

4OA3 Event Follow-up

.1 (Closed) LER 05000255/2006004-00: Reactor Protection System (RPS) Actuation

On May 4, 2006, with the plant in Mode 5, an RPS actuation occurred from the A steam generator low level trip circuitry. Just prior to the occurrence, the secondary side was being maintained at 1.3 psi nitrogen pressure, which was normal chemistry control for existing plant conditions. When the main steam isolation valves (MSIVs) were opened to support a planned test activity, the sudden release of the nitrogen pressure created a momentary steam generator level change in the downcomer region where level is measured. Steam generator indicated level oscillated approximately 15 percent above and below the 40 percent level that existed prior to opening the MSIVs, causing the 28 percent steam generator low level RPS trip setpoint to be exceeded. There was no actual loss of water inventory. Since the RPS was reset and a single control rod was fully withdrawn for testing, the RPS actuation tripped the rod, fully inserting it into the core as expected. No findings were identified. This LER is closed.

.2 (Closed) LER 005000255/2006005-00: Uncoupled Control Rod

On May 10, 2006, an unexpected quadrant power tilt was identified during startup physics testing. At the time of discovery, power was being maintained at approximately 22 percent, following the initial power ascension after a refueling outage. Subsequently, it was determined that Control Rod 33 was fully inserted, not fully withdrawn as indicated, and had been uncoupled from its drive assembly since the refueling outage.

Therefore, the upward mode changes into Mode 2 and Mode 1 made previously were performed in violation of TS. Specifically, TS 3.0.4 was violated because TS 3.1.4.D.1 required all rods to be movable prior to entering Mode 2 and TS 5.4, "Procedures" was violated because the workers had failed to recognize that the rod was not coupled during the refueling outage, as required by procedure RFL-R-11, Coupling Control Rods. The manual RPS actuation made in response to this discovery was also reported in this LER. The self-revealed issue associated with this LER was a violation of NRC requirements, and in Inspection Report 2006004 was determined to be a finding of very low safety significance (NCV 05000255/2006004-03). No additional findings were identified. This LER is closed.

.3 (Open) LER 05000255/2006001-00: Potential Loss of Primary Coolant Makeup

Function for Postulated Fire Scenario On February 14, 2006, during review of 10 CFR 50, Appendix R analysis, the licensee identified a condition that could challenge the ability to maintain the primary coolant makeup function as required by Appendix R, paragraph III G, "Fire Protection of Safe Shutdown Capability." The licensee determined that for a fire in Area 13 that affects all the charging pumps, a spurius short in the affected area could affect the suction isolation for the "B" High Pressure Safety Injection Pump (the remaining credited pump for a fire in that area). The licensee reported the issue as an unanalyzed condition in accordance with 10 CFR 50.72(b)(3)(ii)(B) and took immediate corrective action commensurate with the potential significance of the issue to increase the rounds for a roving fire watch. In addition the licensee created compensatory actions to ensure HPSI pump capability. The actions, which would be used during a fire, include stopping the pump for a spurius start and disabling the control circuitry for the suction valve to ensure the valve would be in an open position. Once the suction valve was put in its safety position, the pump could be operated as needed.

The inspectors reviewed these actions on February 15, 2006, and they were determined to be feasible and reliable. The licensee wrote AR01014772 to evaluate the discrepant condition and to determine the extent of condition. This evaluation is ongoing. The issue is more than minor since it impacted the capability and availability of the needed HPSI pump needed to inject water into the core. This issue is not of high significance since there is reasonable assurance that the core would have remained covered for several hours allowing time to restore functionality to a makeup water source. Final resolution will occur after transition to National Fire Protection Association (NFPA) 805.

Enforcement discretion is being considered based on the September 7, 2006, letter from the NRC authorizing the licensees transition to the code and the applicable section of the enforcement policy, "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)."

.4 (Closed) LER 05000255/2006003-00: Completion of Plant Shutdown required by

Technical Specifications On March 29, 2006, the licensee declared the left train of the emergency core cooling system (ECCS) system inoperable after a valve, CV-3070, failed to stroke during preventive maintenance testing. The valve provides a cooled water supply to the HPSI suction after the ECCS system enter recirculation mode during a loss of coolant accident (LOCA). Since attempts to correct the problem were not successful, the licensee performed a reactor shutdown as required by TS LCO 3.5.2.B.1. The licensee completed the required shutdown within the specified completion times. Subsequent to the shutdown, the licensee determined an improper modification to valves supports caused internal parts to wear to the point that the valve actuator could no longer reposition the valve. The improper modification was a violation of 10 CFR 50, Appendix B, Criterion III and was documented as in Inspection Report 2006006 as NCV 05000255/2006006-02. No additional findings were identified. This LER is closed.

.5 (Open) LER 05000255/2006006-00: Inoperable Containment Due to Containment Air

Cooler through-Wall Flaw On November 1, 2006, the licensee discovered a unisolable SW leak on Containment Air Cooler, VHX-4. The leak was a through-wall leak in American Society of Mechanical Engineers (ASME) Code Class III piping. Due to the location of the failure, the flaw could not be identified and characterized to allow continued use. Therefore, the licensee concluded the closed loop portion of the SW section which services the containment was no longer operable and commenced a shutdown in accordance with TS Action 3.6.1 B for an inoperable containment. The licensee has not categorized the flaw and will not be able to do so until the scheduled outage in 2007 when the cooler will be replaced. The licensee completed the repairs to the cooler, which included blanking off the affected tube bundle to restore containment integrity. No additional findings were identified. This LER will remain open until the licensee completes the assessment of the flaw such that the safety aspects of the condition can be fully evaluated and to determine if a violation of NRC requirements existed.

.6 (Closed) LER 05000255/2006002-00: Main Steam Safety Valves (MSSV) Exceeded Lift

Setpoint Acceptance Criteria On March 26, 2006, main steam safety valve RV-0710 failed in place testing. On March 25, 2006, RV-0707 had failed the same surveillance testing. In both cases, the valves lifted, but above the specified lift setpoint. Technical Specifications stipulate that 23 of 24 relief valves must be operable. The licensee determined that the failure mechanism occurred over time; therefore, the licensee concluded that a condition prohibited by TS existed during plant operation. Per procedure, the licensee adjusted the setpoint of the discrepant valves to restore the valves to TS compliance. The licensee determined that the valves suffered from an initial high lift phenomena.

Therefore, subsequent valves were tested in an offsite facility following plant shutdown using an alternate, but also ASME sanctioned, test methodology. All valves tested offsite passed testing. This licensee-identified finding was a violation of TS 3.7.1. The finding is more than minor because it had a credible impact on safety. The licensees evaluation of the as-found lift pressures demonstrated that no design limits would have been challenged; therefore, the finding screened as having very low safety significance, since there was no loss of safety function. No additional findings were identified. The enforcement aspects of the violation are discussed in Section 4OA7. This LER is closed.

4OA5 Other Activities

.1 (Closed) URI 05000255/2006004-06: Failure of Component on 1-2 EDG Causes

Surveillance Failure The inspectors completed an assessment of the licensees evaluation for URI 05000255/2006004-06 associated with the monthly diesel surveillance test conducted on November 20, 2005, during which a fuel line snubber valve failed and caused a fuel leak. The fuel leak on the 1-2 EDG resulted in aborting the surveillance test. The cause was related to a part which had been installed 28 days earlier. The licensee wrote LER 2005-007-00 on this issue and then retracted it based on their assessment the 1-2 EDG was operable, but degraded. The inspectors reviewed the licensees basis for retraction, and after clarification of numerous items, concluded the licensee had developed an adequate basis for the retraction.

Introduction:

A finding of very low safety significance was self-revealed that was associated with an NCV of 10 CFR Part 50, Appendix B, Criterion VIII, "Identification and Control of Materials, Parts and Components" for failing to have adequate control measures needed to prevent the use of defective parts. Specifically, a fuel leak developed on the 1-2 EDG due to installation of a substandard part that resulted in aborting a surveillance test. The failed part has been replaced and there are no other susceptible parts in the diesel engines onsite. This finding resolves and closes URI 05000255/2006004-06.

Description:

On November 20, 2005, while conducting a monthly surveillance test of the 1-2 EDG, a fuel leak developed on the fuel injector for the 5R cylinder. Licensee personnel determined the fuel leak was significant as well as a potential fire hazard and secured the EDG. Troubleshooting identified that the snubber valve for 5R fuel pump, which had been replaced 28 days earlier, had cracked and caused the fuel oil spray.

The licensee replaced the failed snubber valve and verified operability of the EDG.

Subsequent testing of the snubber indicated that an improper heat treatment had been applied to the snubber valve.

The licensee contacted the EDG vendor that had supplied the snubber valve. The vendor reviewed the data related to the valve and performed additional testing. Based on the result, the vendor confirmed an improper heat treatment had caused the valves failure. However, the vendor could not correlate the obtained data to any snubber they had produced since 1995. The vendor hypothesized that the licensees stock may have been contaminated with stock from a batch produced in the 1990's. Previous operating experience (OE) in 1993 from Diablo Canyon had identified snubbers of improper material. The licensees root cause of the event identified weaknesses in the procurement process that could lead to a loss of part traceability. Therefore, consistent with the licensees root cause, the inspectors concluded that the snubber installed in October of 2005 was substandard due to poor control of procured parts. The vendor stated that incorrectly heat-treated snubbers would fail within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of machine run.

After running the 1-2 EDG for over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, no other failures were discovered. No failures were observed on the other safety-related EDG after running the machine for over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Analysis:

The inspectors concluded that the failure to ensure the proper material from site stock for the safety-related application was a performance deficiency which was within the licensees ability to foresee and correct. The finding is more than minor since the defective part impacted the Mitigating Systems Cornerstone for availability, reliability and capability of the class 1E, onsite EDG system and its associated attribute of equipment performance. The licensee completed an assessment of EDG operability in June of 2006 which concluded the snubber failure did not render the EDG inoperable.

The licensee evaluated the EDG performance due to reduced load capability for one fuel cylinder producing no load. In addition, the licensee evaluated the EDG performance and functional capability from a potential fire hazard due to fuel spray. The inspectors reviewed the licensees assumptions, assessments and follow-on evaluations in detail. Although the licensee determined in late June the diesel could perform its safety function with the failed snubber, the inspectors could not arrive at the same conclusion after reviewing the technical analysis performed by the licensee.

Subsequent interactions between the inspectors and the licensee caused the licensee to more thoroughly evaluate the condition and develop an adequate technical basis to support EDG operability. After numerous interactions with the licensee, the inspectors concluded the expanded basis documents provided adequate support to conclude the EDG was operable but degraded. The licensee wrote CAPs to address the issues raised by the inspectors. Even though the failed snubber did not render the EDG inoperable, the failure resulted in dispersion of flammable material on and near safety-related equipment, reduced the capability of the EDG and required additional unavailability time to correct the failure. Based on this conclusion, the inspectors determined that the issue screened as Green on question 1 (IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations,"

1) since there was no loss of safety function.

Enforcement:

10 CFR Part 50, Appendix B, Criterion VIII, "Identification and Control of Materials, Parts and Components" requires, in part, measures to be established for the identification and control of materials, parts and components. The identification and control measures shall be designed to prevent the use of incorrect parts. Contrary to this, on or about October 24, 2005, the sub-standard part for the 5R fuel line snubber was installed into the 1-2 EDG, a safety-related component, with the wrong heat treatment. As identified in the licensees root cause, historic methods of part control were inadequate to prevent introduction of a defective part. However, because this violation was of very low safety significance and because the issue was entered into the licensee's corrective action program (AR01004766) this violation is being treated as an NCV, consistent with Section VI.A.1 of the Enforcement Policy (NCV 05000255/2006013-04).

.2 Institute of Nuclear Power Operations (INPO) Plant Assessment Report Review

a. Inspection Scope

The inspectors reviewed the final report for the INPO plant assessment of Palisades Nuclear Plant conducted in June 2006. The inspectors reviewed the report to ensure that issues identified were consistent with the NRC perspectives of licensee performance and to verify if any significant issues were identified that required further NRC follow-up.

b. Findings

No findings of significance were identified.

.3 (Closed) URI 07200007/2004002-02: Subsurface Bearing Stability Beneath the

Independent Spent Fuel Storage Installation (ISFSI) Pad During an inspection of pre-operational activities associated with dry fuel storage, the inspectors reviewed the licensees slope stability analyses of the slope where the new ISFSI pad was located. The licensee used acceptance criteria listed in the Department of the Navy NAVFAC DM-7 document, dated March 1971, entitled "Soil Mechanics, Foundations and Earthern Structures." The document recommended a minimum safety factor of 1.15 for transient loads such as earthquakes. The inspectors determined that the licensees initial analyses of dynamic loads under the new pad did not result in a safety factor of 1.15 for all cases (NRC Inspection Report No. 07200007/2004002).

The NRC staff reviewed the licensees slope stability analysis, Calculation No.

EA-EAR-2000-0309-2, Revision 4, dated July 8, 2004, as well as other recognized publications related to the topic. The Department of the Navy document mentioned above, Chapter 7, Section 3.g (4) - Required Safety Factor, stated that, "for transient loads, such as earthquake, safety factors as low as 1.2 or 1.15 may be tolerated." The Naval Facilities Engineering Command re-validated this requirement by Change 1 in September 1986. The commercial standards such as American National Standards Institute, American Society of Civil Engineers, and others also indicate that a minimum acceptable factor of safety should be 1.15 when loadings include transient loadings such as a design basis seismic event. Therefore, upon further review of the issue, the staff found that the licensees rationale for accepting a factor of safety below that established for the design of the ISFSI pad (as low as 1.02 vs. 1.15 minimum) was not consistent with accepted commercial standards and practices; the analysis of record presented in the subject calculation was insufficient; and the licensee needed to revise its analysis accordingly.

Subsequently, the licensee submitted a new calculation for the slope stability analysis, NMC Calculation (Doc) No: EA-EC7408-02, Revision 0, "Re-evaluation of Slope Stability under ISFSI Pad for Revised Load Due to 24PTH System," dated October 19, 2006. The NRC staff verified that the number of soil samples taken in the vicinity of the proposed pad was adequate to determine the soil properties to be used in the design of the pad. The licensees revised evaluation appropriately considered the weight of the as-built pad, the weight of the casks due to the NUHOMS-24PTH system, and the in-situ soil properties in response to a seismic event. The evaluation demonstrated that the design criterion was met for all areas and soils beneath and immediately around the pad.

Based on the review of the assumptions, methods, and conclusions contained in the licensees revised slope stability analysis, the NRC staff concluded that the licensee satisfactorily demonstrated that the as-built pad has a minimum safety factor of 1.15 against the postulated sliding soil mass loads. This URI closed.

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. P. Harden and other members of licensee management on January 10, 2007. 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

Interim exit meetings were conducted for:

• Radioactive Material Processing and Transportation and Reactor Coolant System Specific Activity Performance Indicator with Mr. P. Harden on December 1, 2006.

• ALARA Planning and Controls Program with Mr. P. Harden on December 15, 2006.

• Emergency Preparedness Inspection with Ms. J. Ford on December 21, 2006.

• Independent Spent Fuel Storage Installation with Mr. D. Malone on December 27, 2006.

4OA7 Licensee-Identified Violations

The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.

• The licensee identified a violation of 10 CFR 50.72 on November 4, 2006 when a member of the regulatory affairs staff noted that the site had failed to make a required 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report regarding the loss of the AFW safety function. Upon recognition that 10 CFR 50.72 required a report, the licensee made the appropriate notifications (EN #42963, November 4, 2006). 10 CFR 50.72 requires notification within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of discovery of a condition that could have prevented fulfillment of a safety function; however, the actual notification occurred 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> after discovery. Since the inspectors were on site and discovered the switch mis-positioning that caused all three AFW pumps to be inoperable, the inspectors were aware of the condition. The licensee promptly restored AFW to operability upon discovery. Since failure to report events in accordance with 10 CFR 50.72 affects the NRCs ability to perform its regulatory function, the inspectors processed the violation as traditional enforcement consistent with NRC Enforcement Policy IV.A.3 and the Enforcement Manual.

The licensee wrote CAP01060168, Late Determination of 8-hour Emergency Report, November 8, 2006, to address the issue. Since the licensee continued to review reportability until they ultimately determined the inoperability of the AFW pumps was reportable, the inspectors concluded the violation was licensee identified and not more than very low safety significance.

• Technical Specification 3.7.1 requires a minimum of 23 of the 24 MSSVs to be operable in modes 1, 2 and 3. Contrary to this on March 26, 2006, with the plant in mode 1, the licensee determined during surveillance testing that two MSSVs were inoperable during power operations with lift setpoints outside of TS table 3.7.1-1. The licensee wrote CAP 01020547, Main Steam Safety Valves, on March 26, 2006. The licensee-identified violation is of very low safety significance since there was no actual loss of safety function for the steam relief system with the setpoints slightly out of tolerance.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

P. Harden, Site Vice President

S. Bell, Senior Health Physicist

B. Berles, System Engineering Manager

T. Blake, Nuclear Safety Assurance Manager

A. Blind, Design Engineering Manager

L. Blocker, Operations Manager

J. Broschak, Engineering Director

B. Dotson, Regulatory Compliance

G. Baustian, Training Manager

W. Edwards, Chemistry Technician

J. Ford, Emergency Preparedness Manager

G. Hettel, Plant Manager

G. Higgs, Maintenance Manager

P. Johnson, Safety Manager

C. Jones, Chemist

L. Lahti, Licensing Manager

A. Lyon, Design Engineer

D. Malone, Regulatory Affairs

C. Moeller, Radiation Protection General Supervisor

D. Nestle, Radiation Protection General Supervisor - Technical

B. Nixon, Work Control Manager

B. Patrick, Radiation Protection Manager

G. Sleeper, Assistant Operations Manager

G. Sturm, ALARA Specialist

D. Watkins, Radwaste Shipping Analyst

P. Williams, Sr. RP Technician - Outage ALARA Planner

K. Yeager, Assistant Operations Manager

Nuclear Regulatory Commission

M. Chawla, Project Manager, NRR

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000255/2006013-01 NCV Inaccurate Surveillance Procedure for RCS Leakrate Calculation (Section 1R22)

05000255/2006013-02 FIN Failure to Adequately Implement Radiological Dose Controls (Section 2OS2)

Attachment

05000255/2006013-03 URI Potential Impact on AFW Pumps to High Energy Line Breaks in the Turbine Building (Section 4OA2)

05000255/2006013-04 NCV Failure of Component on 1-2 EDG Causes Surveillance Failure (Section 4OA5)

Closed

05000255/2006013-01 NCV Inaccurate Surveillance Procedure for RCS Leakrate Calculation (Section 1R22)

05000255/2006013-02 FIN Failure to Adequately Implement Radiological Dose Controls (Section 2OS2)

05000255/2006004-00 LER Reactor Protection System Actuation (Section 4OA3)

05000255/2006005-00 LER Uncoupled Control Rod (Section 4OA3)

05000255/2006003-00 LER Completion of Plant Shutdown Required By Technical Specifications (4OA3)

05000255/2006002-00 LER Main Steam Safety Valves Exceeded Lift Setpoint Acceptance Criteria (Section 4OA3)

05000255/2006004-06 URI Failure of Component on 1-2 EDG Causes Surveillance Failure (Section 4OA5)

05000255/2006013-04 NCV Failure of Component on 1-2 EDG Causes Surveillance Failure (Section 4OA5)

200007/2004002-02 URI Subsurface Bearing Stability Beneath the ISFSI pad (Section 4OA5)

Discussed

05000255/2006001-00 LER Potential Loss of Primary Coolant Makeup Function for Postulated Fire Scenario (Section 4OA3)

05000255/2006006-00 LER Inoperable Containment Due to Containment Air Cooler through-Wall Flaw (Section 4OA3)

Attachment