IR 05000255/2011002

IR 05000255-11-002; 01/01/2011 - 03/31/2011; Palisades Nuclear Plant; Operability Evaluations, Identification and Resolution of Problems, Followup of Events and Notices of Enforcement Discretion, Other Activities

ML111240114
Person / Time
Site:	Palisades, Rensselaer Polytechnic Institute
Issue date:	05/04/2011
From:	Jack Giessner Reactor Projects Region 3 Branch 4
To:	Kirwin T Entergy Nuclear Operations
References
IR-11-002
Download: ML111240114 (41)

Text

UNITED STATES May 4, 2011

SUBJECT:

PALISADES NUCLEAR PLANT INTEGRATED INSPECTION REPORT 05000255/2011002

Dear Mr. Kirwin:

On March 31, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Palisades Nuclear Plant. The enclosed report documents the results of this inspection, which were discussed on April 13, 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, three NRC-identified findings and one self-revealed finding of very low safety significance were identified. Three of the findings involved a violation of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your corrective action program, the NRC is treating the issues as non-cited violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy. Additionally, a licensee-identified violation is listed in Section 4OA7 of this report.

If you contest the subject or severity of an NCV, 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, 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 Resident Inspector Office at the Palisades Nuclear Plant. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Palisades Nuclear Plant. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS)

component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

John B. Giessner, Chief Branch 4 Division of Reactor Projects Docket No. 50-255 License No. DPR-20

Enclosure:

Inspection Report 05000255/2011002; w/Attachment: Supplemental Information

REGION III==

Docket No: 50-255 License No: DPR-20 Report No: 05000255/2011002 Licensee: Entergy Nuclear Operations, Inc.

Facility: Palisades Nuclear Plant Location: Covert, MI Dates: January 1, 2011, to March 31, 2011 Inspectors: J. Ellegood, Senior Resident Inspector T. Taylor, Resident Inspector J. Cassidy, Senior Health Physicist Approved by: John B. Giessner, Chief Branch 4 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000255/2011002; 01/01/2011 - 03/31/2011; Palisades Nuclear Plant; Operability

Evaluations, Identification and Resolution of Problems, Followup of Events and Notices of Enforcement Discretion, Other Activities This report covers a 3-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. It includes four green findings, three of which were considered non-cited violations (NCV) of Nuclear Regulatory (NRC) regulations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP).

Cross-cutting aspects were determined using IMC 0310, Components Within the Cross-Cutting Areas. 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 4, dated December 2006.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Initiating Events

Green.

A finding of very low safety significance without an associated NCV was self-revealed when a loss of the rear bus and loss of one cooling tower occurred.

The licensee failed to maintain the enclosure for F and G busses weatherproof as stipulated in the design basis documents for the 4160V electrical system. In addition, the licensee cancelled a preventive maintenance task to inspect the enclosures caulking. Due to degradation of the seals, water intruded into the F bus switchgear and caused a short and explosion resulting in loss of one qualified circuit of offsite power. This resulted in entry into an Emergency Action Level (EAL) of an Usual Event (the lowest emergency classification). As an immediate action, the licensee reduced power to about 55 percent. The licensee entered the finding into their corrective action program (CAP).

The finding was more than minor because it impacted the initiating event cornerstone objective of limiting the likelihood of those events that upset plant stability and is associated with the attribute of equipment performance. Using IMC 0609 Appendix A the inspectors determined the finding was of very low safety significance because even though the issue impacted the transient initiating event frequency, it did not impact the mitigating system availability. The inspectors determined there was no cross-cutting aspect because the causes of the failure to maintain the switchgear enclosure are not reflective of current performance. There was no violation of NRC requirements. (4OA3)

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to recognize and account for potential age-related degradation of capacitors in the emergency diesel generator (EDG) digital reference units design controls.

Specifically, the installed capacitors were found beyond industry and vendor recommended useful life and if they were to degrade, could impact safety-related functions of the EDGs. The licensee entered the issue into their CAP and replaced the digital reference units.

The issue was more than minor because if left uncorrected, it could become a more significant safety concern because the capacitors would continue to degrade. The finding affected the Mitigating Systems Cornerstone and screened as very low safety significance (Green) based on the assessment that the operability of the EDG was maintained, and answering no to all questions for that cornerstone in IMC 0609 Attachment 4, table 4a. The finding had an associated cross-cutting aspect in the area of Problem Identification and Resolution. Specifically, the licensee did not use operating experience information, including vendor recommendations, to support plant safety in that relevant information was not collected, evaluated, and communicated in a timely manner. Although the part 21 was issued in 2001, the licensee had the opportunity to identify the condition in March 2011 when evaluating the acceptability for continued use of EDG governor components that were also impacted by the 2001 part 21. (P.2(a)) (1R15)

Cornerstone: Barrier Integrity

Green.

The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 26.205(d) for the failure to control the work hours of covered workers. Specifically, contract workers violated the minimum days off requirements during the October 2010 refueling outage and were not being tracked and controlled in accordance with licensee procedures. The licensee entered the issue into their CAP and reviewed the hours worked and jobs performed by the contract workers.

The issue affected the Barrier Cornerstone because the work being performed involved reactor fuel and was more than minor because if left uncorrected, it could become a more significant safety concern. The finding screened as very low safety significance (Green) based on no known effects to the plant caused by possible worker fatigue. The finding had an associated cross-cutting aspect in the human performance area. Specifically, the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee did not ensure work hours were tracked appropriately for personnel doing covered work (H.4(c)). (4OA2)

Cornerstone: Public Radiation Safety

Green.

The inspectors identified a finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4 for failure to establish and implement procedures recommended by Regulatory guide 1.33. Specifically, the licensee failed to establish procedures for liquid radioactive waste and emergency procedures for abnormal releases of radioactivity related to tank T-90 and 91. The licensee has revised procedures to control concentrations of tritium in tanks T-90 and 91 and entered the condition into the CAP.

The inspectors concluded that the failure to maintain procedures as required by TS 5.4 was a performance deficiency that warranted a significance determination.

The inspectors determined the finding was more than minor because it impacted the public radiation safety cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation, in that, the licensee failed to meet the program and process attribute of procedures. Since the finding resulted in less than .005 rem exposure to members of the public, the inspectors concluded the finding was of very low safety significance (green) in accordance with IMC 0609, Appendix D. There was no cross-cutting aspect in that the procedures and Updated Final Safety Analysis Review (UFSAR) content have been in place for several years and do not reflect current plant performance. (4OA5)

Licensee-Identified Violations

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

REPORT DETAILS

Summary of Plant Status

The Plant began the inspection period at 100 percent power. On January 8, a fault in switchgear providing power to the forced draft cooling towers resulted in loss of a cooling tower and a downpower to about 55 percent power. After repair of the switchgear, the licensee ascended to 100 percent power on January 16. On January 22, the plant tripped due to a fault in the cabling to the G bus. The licensee realigned the electrical distribution system and restarted the plant on January 24. On January 26, the licensee reached 100 percent power and remained at or near 100 percent for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

a. Inspection Scope

On February 2, 2011, a blizzard warning was issued for expected blizzard conditions.

The inspectors observed the licensees preparations and planning for the significant winter weather potential. The inspectors reviewed licensee procedures and discussed potential compensatory measures with control room personnel. The inspectors focused on plant managements actions for implementing the stations procedures for ensuring adequate personnel for safe plant operation and emergency response would be available. The inspectors conducted a site walkdown including walkdowns of various plant structures and systems to check for maintenance or other apparent deficiencies that could affect system operations during the predicted significant weather. The inspectors also reviewed corrective action program (CAP) items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their CAP in accordance with station corrective action procedures.

Specific documents reviewed during this inspection are listed in the Attachment.

This inspection constituted one readiness for impending adverse weather condition sample as defined in Inspection Procedure (IP) 71111.01-05.

b. Findings

No findings were identified.

1R04 Equipment Alignment

a. Inspection Scope

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

• Left train control room heating, ventilation and air-conditioning with the right train out-of-service;

• auxiliary feedwater;

• 1-1 emergency diesel generator (EDG) with 1-2 EDG out-of-service; and

• service water with service water C pump out of service.

The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, UFSAR, Technical Specification (TS) requirements, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the CAP with the appropriate significance characterization. Documents reviewed are listed in the to this report.

These activities constituted four partial system walkdown samples as defined in IP 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

a. Inspection Scope

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

• Fire area 16, component cooling water pump room;

• Fire area 24, auxiliary feedwater pump room;

• Fire areas 11, battery room 2;

• Fire area 12, battery room 1; and

• Fire area 14, containment 590.

The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and implemented adequate compensatory measures for out-of-service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensees fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event. Using the documents listed in the Attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees CAP. Documents reviewed are listed in the Attachment to this report.

These activities constituted five quarterly fire protection inspection samples as defined in IP 71111.05-05.

b. Findings

No findings were identified.

1R07 Annual Heat Sink Performance

a. Inspection Scope

The inspectors observed the licensees inspection of diesel generator 1-2 heat exchangers to verify that potential deficiencies did not impede the heat exchangers ability to remove heat. The inspectors looked for evidence of blocked tubes and tube fouling. In addition the inspectors looked at the service water supply piping to the heat exchangers to verify that interior corrosion did not impact the supply of service water to the heat exchangers. The inspectors reviewed the licensees inspection checklist and eddy current test results to verify that the diesel was operable. Documents reviewed for this inspection are listed in the Attachment to this document.

This annual heat sink performance inspection constituted one sample as defined in IP 71111.07-05.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

a. Inspection Scope

On February 9, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator requalification examinations to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• licensed operator performance;

• crews clarity and formality of communications;

• ability to take timely actions in the conservative direction;

• prioritization, interpretation, and verification of annunciator alarms;

• correct use and implementation of abnormal and emergency procedures;

• control board manipulations;

• oversight and direction from supervisors; and

• ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one quarterly licensed operator requalification program sample as defined in IP 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk significant systems:

• 4160V electrical distribution; and

• reactor protection system.

The inspectors reviewed events such as where ineffective equipment maintenance had resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• implementing appropriate work practices;

• identifying and addressing common cause failures;

• scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

• characterizing system reliability issues for performance;

• charging unavailability for performance;

• trending key parameters for condition monitoring;

• ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and

• verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2), or appropriate and adequate goals and corrective actions for systems classified as (a)(1).

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

This inspection constituted two quarterly maintenance effectiveness samples as defined in IP 71111.12-05.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

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

• emergent work to restore F bus;

• forced outage activities and radiation monitor failures;

• blizzard warnings prior to planned diesel outage;

• fire pump maintenance; and

• EDG 1-1 outage.

These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.

These maintenance risk assessments and emergent work control activities constituted five samples as defined in IP 71111.13-05.

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• containment sump due to transient material;

• cracks in EDG turbocharger support plates;

• pressurizer reliefs with inadequate VT-2 exam;

• deferred replacement of EDG governors, and;

• 4160V bus issues.

The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. 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 compared the operability and design criteria in the appropriate sections of the TS and UFSAR to the licensees evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the to this report.

This operability inspection constituted five samples as defined in IP 71111.15-05.

b. Findings

Introduction.

The inspectors identified a finding of very low safety significance (Green)and associated non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion III, Design Control, for the failure to account for the effects of aging on certain electronic components in the EDG governor systems. Specifically, electrolytic capacitors in the digital reference unit were found beyond the vendor and industry recommended replacement interval.

Description.

During a recent planned maintenance outage for one of the sites EDGs, the licensee decided to defer replacement of components in the governor system. The licensee decided to defer the replacement, because the time to complete the work could challenge the remaining time in the TS allowed outage window. The inspectors questioned the licensee as to why this would be acceptable given that one of the electronic components, designated as the EGA, had an industry recommended refurbishment interval of approximately 7-10 years. The installed EGA had been in place for approximately 16 years. The components of concern were electrolytic capacitors that degrade over time and affect the operation of the EGA. This and other governor issues had been highlighted in a Part 21 report from 2001. Additionally, some industry guidance had been published in the form of an Electric Power Research Institute report and ALCO Engine Owners Group report, both of which described recommended maintenance practices for EDGs. The ALCO report specifically addressed electrolytic capacitors in governor systems. After a few days of research, the licensee validated that the capacitors in the EGA had been replaced years ago with longer-life tantalum capacitors, which alleviated the concern for age-related effects.

However, during their review of the issue, the inspectors identified that capacitors in another component, the digital reference unit, could also suffer from age-related degradation. The same industry documents referenced by the licensee to evaluate the issue with the EGAs, also listed the digital reference units as having susceptible capacitors. Since installation in 1998, the licensee had not recognized that the capacitors in the digital reference unit were susceptible to age-related degradation and that the vendor recommended refurbishment interval was approximately 10 years.

When the inspectors raised the issue, the digital reference unit capacitors were approximately 4 years beyond the industry recommended refurbishment interval.

Degradation of the capacitors in the digital reference unit could impact safety-related functions of the EDGs in that the output of the digital reference unit could be affected, which could impact frequency control. The licensee reviewed other industry research and consulted a vendor. Based on Palisades-specific operating conditions, the licensee performed an operability evaluation and concluded the useful life of the capacitors could be extended from the published life until December 2011. The licensee entered the issue into their CAP and has since replaced both digital reference units.

Analysis.

The failure to account for potential degradation of components in the design of safety-related equipment was a performance deficiency warranting further evaluation with the SDP. Specifically, the licensee did not recognize that certain capacitors were susceptible to age-related degradation that could impact safety the function of the EDGs.

In accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, the issue was determined to be more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the capacitors would reach the end of useful life in December 2011 and the licensee had no plans to replace the capacitors without inspector intervention. Inspection Manual Chapter 0612, Appendix E, examples 3j and 3k, support the inspectors conclusion that the finding is more than minor. The finding affected the Mitigating Systems Cornerstone. Using IMC 0609, Appendix A, for a finding during at-power situations and Attachment 4, the inspectors determined the issue screened as Green, or very low safety significance, based on answering no to the questions under the Mitigating Systems Cornerstone in Table 4a.

Specifically, the condition did not affect current or past operability of the EDGs.

The finding had an associated cross-cutting aspect in the area of Problem Identification and Resolution. Specifically, the licensee did not use operating experience information, including vendor recommendations, to support plant safety in that relevant information was not collected, evaluated, and communicated in a timely manner (P.2.(a)). Although the applicable operating experience was in 2001, the licensee failed to identify the concern when researching that acceptability of the capacitors in the EGAs in 2011 during an EDG outage.

Enforcement.

10 CFR 50, Appendix B, Criterion III, Design Control, required, in part, that the licensee shall establish measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of structures, systems, and components. Contrary to this, from their installation in 1998 through March 18, 2011, the licensee did not have a process to ensure that capacitors in the digital reference unit remained suitable for installation based on their recommended useful life. Additionally, the licensee missed an opportunity to identify that digital reference unit components in the Palisades EDGs were susceptible to age-related degradation in 2001 when Part 21 reports were evaluated. The licensee entered the issue into their CAP as CR-PLP-2011-01341 and has since replaced the digital reference unit in each EDG. Because this violation was of very low safety significance and it was entered into the licensees CAP, this violation is being treated as an NCV, consistent with the Enforcement Policy:

NCV 05000255/2011002-01, Failure to Account for Potential Age-Related Degradation in EDG Governors

1R18 Plant Modifications

a. Inspection Scope

The inspectors reviewed the following temporary modification(s):

• temporary shelter around F and G busses.

The inspectors compared the temporary configuration changes and associated 10 CFR 50.59 screening and evaluation information against the design basis, the UFSAR, and the TS, as applicable, to verify that the modification did not affect the operability or availability of the affected system(s). The inspectors also compared the licensees information to operating experience information to ensure that lessons learned from other utilities had been incorporated into the licensees decision to implement the temporary modification. The inspectors, as applicable, performed field verifications to ensure that the modifications were installed as directed; the modifications operated as expected; modification testing adequately demonstrated continued system operability, availability, and reliability; and that operation of the modifications did not impact the operability of any interfacing systems. Lastly, the inspectors discussed the temporary modification with operations, engineering, and training personnel to ensure that the individuals were aware of how extended operation with the temporary modification in place could impact overall plant performance. Documents reviewed are listed in the to this report.

This inspection constituted one temporary modification sample as defined in IP 71111.18-05.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• F bus following repair;

• P-7A, A service water pump following motor replacement;

• EDG 1-1 following outage;

• diesel fire pump following maintenance outage;

• radiation monitors following modification;

• C auxiliary feedwater pump following damage to local switch; and

• replacement of contactor in reactor protection system These activities were selected based upon the structure, system, or component's ability to impact risk. The inspectors evaluated these activities for the following (as applicable):

the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion); and test documentation was properly evaluated. The inspectors evaluated the activities against TS, the Updated Final Safety Analysis Report (UFSAR), 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with post-maintenance tests to determine whether the licensee was identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety. Documents reviewed are listed in the Attachment to this report.

This inspection constituted seven post-maintenance testing samples as defined in IP 71111.19-05.

b. Findings

No findings were identified.

1R20 Outage Activities

a. Inspection Scope

The inspectors evaluated outage activities for a forced outage that began on January 22 and continued through the January 24, 2011. The inspectors reviewed activities to ensure that the licensee considered risk in developing, planning, and implementing the outage schedule.

The inspectors observed control room response to maintain the plant stable following the reactor trip. During the forced outage, the inspectors observed outage equipment configuration and risk management, electrical lineups, selected clearances, control and monitoring of decay heat removal, control of containment activities, startup activities, and identification and resolution of problems associated with the outage. The outage occurred as a result of a plant trip due to a turbine trip. The licensee determined the turbine tripped as a result of a fault on the Y phase between the 1-3 station power transformer and the G bus. The licensee isolated the faulted cable but has not determined why the cable failed.

This inspection constituted one other outage sample as defined in IP 71111.20-05.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

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

• MO-7A-1, EDG monthly test (routine);

• T-265, Test of EDG breaker 152-213 trips (routine);

• DWO-1, power monitoring as part of daily/weekly checks (routine) ; and

• A component cooling water pump (inservice).

The inspectors observed in-plant activities and reviewed procedures and associated records to determine the following:

• did preconditioning occur;

• were the effects of the testing adequately addressed by control room personnel or engineers prior to the commencement of the testing;

• were acceptance criteria clearly stated, demonstrated operational readiness, and consistent with the system design basis;

• plant equipment calibration was correct, accurate, and properly documented;

• as-left setpoints were within required ranges; and the calibration frequency was in accordance with TSs, the USAR, procedures, and applicable commitments;

• measuring and test equipment calibration was current;

• test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied;

• test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used;

• test data and results were accurate, complete, within limits, and valid;

• test equipment was removed after testing;

• where applicable for inservice testing activities, testing was performed in accordance with the applicable version of Section XI, American Society of Mechanical Engineers code, and reference values were consistent with the system design basis;

• where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable;

• where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure;

• where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished;

• prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test;

• equipment was returned to a position or status required to support the performance of its safety functions; and

• all problems identified during the testing were appropriately documented and dispositioned in the CAP.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted three routine surveillance testing samples and one inservice testing sample, as defined in IP 71111.22, Sections -02 and -05.

b. Findings

No findings were identified.

1EP6 Drill Evaluation

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of a routine licensee emergency drill on February 23, 2011, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the simulator control room, technical support center, and emergency operations facility to determine whether the event classification, notifications, and protective action recommendations were performed in accordance with procedures. The inspectors also attended the licensee drill critique to compare any inspector-observed weakness with those identified by the licensee staff in order to evaluate the critique and to verify whether the licensee staff was properly identifying weaknesses and entering them into the CAP. As part of the inspection, the inspectors reviewed the drill package and other documents listed in the to this report.

This emergency preparedness drill inspection constituted one sample as defined in IP 71114.06-05.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation

This inspection constituted one complete sample as defined in IP 71124.05-05.

.1 Inspection Planning (02.01)

a. Inspection Scope

The inspectors reviewed the plant UFSAR to identify radiation instruments associated with monitoring area radiological conditions including airborne radioactivity, process streams, effluents, materials/articles, and workers. Additionally, the inspectors reviewed the instrumentation and the associated TS requirements for post-accident monitoring instrumentation including instruments used for remote emergency assessment.

The inspectors reviewed a listing of in-service survey instrumentation including air samplers and small article monitors, along with instruments used to detect and analyze workers external contamination. Additionally, the inspectors reviewed personnel contamination monitors and portal monitors including whole-body counters to detect workers internal contamination. The inspectors reviewed this list to assess whether an adequate number and type of instruments are available to support operations.

The inspectors reviewed licensee and third-party evaluation reports of the radiation monitoring program since the last inspection. These reports were reviewed for insights into the licensees program and to aid in selecting areas for review (smart sampling).

The inspectors reviewed procedures that govern instrument source checks and calibrations, focusing on instruments used for monitoring transient high radiological conditions, including instruments used for underwater surveys. The inspectors reviewed the calibration and source check procedures for adequacy and as an aid to perform smart sampling.

The inspectors reviewed the area radiation monitor alarm setpoint values and setpoint bases as provided in the TS and the UFSAR.

The inspectors reviewed effluent monitor alarm setpoint bases and the calculational methods provided in the Offsite Dose Calculation Manual (ODCM).

b. Findings

No findings were identified.

.2 Walkdowns and Observations (02.02)

a. Inspection Scope

The inspectors walked down effluent radiation monitoring systems, including at least one liquid and one airborne system. Focus was placed on flow measurement devices and all accessible point-of-discharge liquid and gaseous effluent monitors of the selected systems. The inspectors assessed whether the effluent/process monitor configurations align with ODCM descriptions and observed monitors for degradation and out-of-service tags.

The inspectors selected portable survey instruments in use or available for issuance and assessed calibration and source check stickers for currency as well as instrument material condition and operability.

The inspectors observed licensee staff performance as the staff demonstrated source checks for various types of portable survey instruments. The inspectors assessed whether high-range instruments are source checked on all appropriate scales.

The inspectors walked down area radiation monitors and continuous air monitors to determine whether the monitors were appropriately positioned relative to the radiation sources or areas they were intended to monitor. Selectively, the inspectors compared monitor response (via local or remote control room indications) with actual area conditions for consistency.

The inspectors selected personnel contamination monitors, portal monitors, and small article monitors and evaluated whether the periodic source checks were performed in accordance with the manufacturers recommendations and licensees procedures.

b. Findings

No findings were identified.

.3 Calibration and Testing Program (02.03)

Process and Effluent Monitors

a. Inspection Scope

The inspectors selected effluent monitor instruments (such as gaseous and liquid) and evaluated whether channel calibration and functional tests were performed consistent with radiological effluent TS/ODCM. The inspectors assessed whether:

(a) the licensee calibrated its monitors with National Institute of Standards and Technology traceable sources;

(b) the primary calibrations adequately represented the plant nuclide mix;

(c) when secondary calibration sources were used, the sources were verified by the primary calibrations; and

(d) the licensees channel calibrations encompassed the instruments alarm set-points.

The inspectors assessed whether the effluent monitor alarm setpoints are established as provided in the ODCM and station procedures.

For changes to effluent monitor setpoints, the inspectors evaluated the basis for changes to ensure that an adequate justification exists.

b. Findings

No findings were identified.

Laboratory Instrumentation

a. Inspection Scope

The inspectors assessed laboratory analytical instruments used for radiological analyses to determine whether daily performance checks and calibration data indicate that the frequency of the calibrations is adequate and there were no indications of degraded instrument performance.

The inspectors assessed whether appropriate corrective actions were implemented in response to indications of degraded instrument performance.

b. Findings

No findings were identified.

Whole Body Counter

a. Inspection Scope

The inspectors reviewed the methods and sources used to perform whole body count functional checks before daily use of the instrument and assessed whether check sources were appropriate and align with the plants isotopic mix.

The inspectors reviewed whole body count calibration records since the last inspection and evaluated whether calibration sources were representative of the plants source term and that appropriate calibration phantoms were used. The inspectors looked for anomalous results or other indications of instrument performance problems.

b. Findings

No findings were identified.

Post-Accident Monitoring Instrumentation

a. Inspection Scope

Inspectors selected containment high-range monitors and reviewed the calibration documentation since the last inspection.

The inspector assessed whether an electronic calibration was completed for all decades above 10 rem/hour and whether at least one decade at or below 10 rem/hour was calibrated using an appropriate radiation source.

The inspectors assessed whether calibration acceptance criteria are reasonable, accounting for the large measuring range and the intended purpose of the instruments.

The inspectors selected two effluent/process monitors that are relied on by the licensee in its emergency operating procedures as a basis for triggering emergency action levels and subsequent emergency classifications, or to make protective action recommendations during an accident. The inspectors evaluated the calibration and availability of these instruments.

The inspectors reviewed the licensees capability to collect high-range, post accident iodine effluent samples.

As available, the inspectors observed electronic and radiation calibration of these instruments to verify conformity with the licensees calibration and test protocols.

b. Findings

No findings were identified.

Portal Monitors, Personnel Contamination Monitors, and Small Article Monitors

a. Inspection Scope

For each type of these instruments used on site, the inspectors assessed whether the alarm setpoint values are reasonable under the circumstances to ensure that licensed material is not released from the site.

The inspectors reviewed the calibration documentation for each instrument selected and discussed the calibration methods with the licensee to determine consistency with the manufacturers recommendations.

b. Findings

No findings were identified.

Portable Survey Instruments, Area Radiation Monitors, Electronic Dosimetry, and Air Samplers/Continuous Air Monitors

a. Inspection Scope

The inspectors reviewed calibration documentation for at least one of each type of instrument. For portable survey instruments and area radiation monitors, the inspectors reviewed detector measurement geometry and calibration methods and had the licensee demonstrate use of its instrument calibrator as applicable. The inspectors conducted comparison of instrument readings versus an NRC survey instrument if problems were suspected.

As available, the inspectors selected portable survey instruments that did not meet acceptance criteria during calibration or source checks to assess whether the licensee had taken appropriate corrective action for instruments found significantly out of calibration (greater than 50 percent). The inspectors evaluated whether the licensee had evaluated the possible consequences of instrument use since the last successful calibration or source check.

b. Findings

No findings were identified.

Instrument Calibrator

a. Inspection Scope

As applicable, the inspectors reviewed the current output values for the licensees portable survey and area radiation monitor instrument calibrator unit(s). The inspectors assessed whether the licensee periodically measures calibrator output over the range of the instruments used through measurements by ion chamber/electrometer.

The inspectors assessed whether the measuring devices had been calibrated by a facility using National Institute of Standards and Technology traceable sources and whether corrective factors for these measuring devices were properly applied by the licensee in its output verification.

b. Findings

No findings were identified.

Calibration and Check Sources

a. Inspection Scope

The inspectors reviewed the licensees 10 CFR Part 61, Licensing Requirements for Land Disposal of Radioactive Waste, source term to assess whether calibration sources used were representative of the types and energies of radiation encountered in the plant.

b. Findings

No findings were identified.

.4 Problem Identification and Resolution (02.04)

a. Inspection Scope

The inspectors evaluated whether problems associated with radiation monitoring instrumentation were being identified by the licensee at an appropriate threshold and were properly addressed for resolution in the licensee CAP. The inspectors assessed the appropriateness of the corrective actions for a selected sample of problems documented by the licensee that involve radiation monitoring instrumentation.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

.1 Routine Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Attributes reviewed included: identification of the problem was complete and accurate; timeliness was commensurate with the safety significance; evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent-of-condition reviews, and previous occurrences reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue. Minor issues entered into the licensees CAP as a result of the inspectors observations are included in the Attachment to this report.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for followup, the inspectors performed a daily screening of items entered into the licensees CAP. This review was accomplished through inspection of the stations daily condition report packages.

These daily reviews were performed by procedure as part of the inspectors daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings were identified.

.3 Selected Issue Followup Inspection: Control of Work Hours

a. Inspection Scope

During a review of items entered in the licensees CAP as part of a review of outage work hours, the inspectors identified an issue that was inappropriately evaluated by the licensee dealing with whether or not a violation of the fatigue rule had occurred. The inspectors raised their concerns with the licensee and reviewed their subsequent evaluation.

This review constituted one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

Introduction.

The inspectors identified a finding of very low safety significance (Green)with an associated NCV of 10 CFR 26.205(d) for the failure to control the work hours of covered workers. Specifically, contract workers violated the minimum days off requirements during the October 2010 refueling outage and were not being tracked in accordance with licensee procedures.

Description.

During a review of hours worked by maintenance personnel during the refueling outage, the inspectors noticed a condition report that stated some contractor personnel in the reactor services group had violated the minimum days off requirements of the fatigue rule. This involved seven workers performing fuel inspection and reconstitution activities. Additionally, it appeared that the allowed number of working hours in a seven day period had also been violated. The condition report disposition concluded that there was no violation of any requirements. The inspectors questioned this conclusion and determined that there actually was a violation of the minimum days off requirements. The licensee had erroneously concluded that since the workers had a 34 hour3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> break in a 9 day period that the minimum days off requirements were satisfied.

In fact, break periods and minimum days off requirements were separate were requirements. After the inspectors raised the issue with the licensee, another condition report was generated which later agreed with the inspectors that a violation of minimum days off had occurred. Discussions between the licensee and inspectors also revealed that the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a 7 day period requirements were met. However, to determine what hours were actually worked by the contract personnel to validate that, the licensee had to rely on interviews and turnstile times because the hours worked were not being updated in the licensees tracking software as required by EN-OM-123, Fatigue Management Program.

The sites condition review group reviewed the second condition report (CR) and determined that the site would perform a higher-tier apparent cause evaluation (ACE).

In the ACE, the licensee concluded that in 2010, there were 20 instances where violations had occurred without condition reports being generated and 35 instances when fatigue tracking records had not been initiated when required, contrary to the requirements of procedure EN-OM-123. The ACE also concluded there were inconsistencies in how condition review group was classifying fatigue rule issues and less than adequate management oversight of the program.

Analysis.

The failure to adequately implement fatigue rule requirements was a performance deficiency warranting further evaluation in the SDP. The issue affected the Barrier Cornerstone because the work being performed involved reactor fuel. The issue was determined to be more than minor per IMC 0612, Appendix B, because if left uncorrected, the issue could become a more significant safety concern. Specifically, the number a violations indicated that a programmatic issue existed. Continued programmatic weaknesses in fatigue management could result in human performance errors that could impact plant equipment. Inspection Manual Chapter 0612, Appendix E, example 9a applies as a benchmark for significance determination as an example which shows the determination to be more than minor. The inspectors used IMC 0609 Appendix G to determine the significance of the finding, because the issue was discovered during the refueling outage and a majority of the issues uncovered by the ACE occurred during the outage. Because there were no known plant issues caused by worker fatigue, the IMC 0609, Appendix G, Attachment 1, assessment determined there was no impacted equipment, therefore a quantitative analysis and Phase 2 and 3 analyses were not required. Therefore, the issue screened as Green, or very low safety significance.

The inspectors reviewed the ACE and determined finding had an associated cross-cutting aspect in the human performance area. Specifically, the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee did not ensure work hours were tracked appropriately for personnel doing covered work (H.4.(c)).

Enforcement.

10 CFR 26.205(d) requires that the licensee shall control the work hours of individuals who are subject to the section. EN-OM-123, the licensees procedure for managing the fatigue rule, implements 10 CFR 26.205(d) requirements. Contrary to both, on October 17, 2010, workers classified as performing covered work under the rule violated the minimum days off requirements by working a seventh consecutive day, and therefore the licensee failed to control their work hours. This violation was not recognized by the licensee until prompted by inspectors. The licensee evaluated the issue under condition reports CR-PLP-2010-05185, 06623, and 2011-00093. Because this violation was of very low safety significance and it was entered into the licensees CAP, this violation is being treated as a NCV, consistent with the Enforcement Policy.

NCV 05000255/2011002-02, Violation of Fatigue Rule Requirements.

.4 Selected Issue Followup Inspection: Operation Above Licensed Power Level:

a. Inspection Scope

On January 18 and into January 19, the licensee operated with an indicated power in excess of the licensed power limit. The inspectors reviewed the licensees actions and procedure for controlling reactor power and concluded the licensee incorrectly determined compliance with the licensed power could be based solely on the 2-hour power average. By procedure, the licensee did not consider other indications of power in excess of the licensed power to require prompt response to reduce power to the licensed power level.

The licensee uses a heat balance calculation performed by the plant process computer to determine the heat generation of the reactor and uses a 2-hour average to determine compliance with the licensed power level. For Palisades, a calculated power of 100.1 percent corresponds to the licensed thermal power level. The licensee uses a computer program to calculate power based on a heat balance. The program calculates a 2-hour average as an average of heat balance steady state (HB_PWR_STEADY) over a rolling 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period. The program calculates this heat balance steady state as an average of seven minutes worth of heat balance calculations performed at 10 second intervals. The licensee uses averages, in part, to dampen variations in indicated power.

RIS-2007-21 recognizes that variations in power occur due to normal fluctuations of plant parameters. However, the RIS also emphasizes that there is no regulatory guidance that condones operation above the licensed power level. Therefore, when power is over the licensed power level, licensees must take prompt action to reduce thermal power. In reviewing the condition, the inspectors noted that the calculation for adjusting the ultrasonic flow measurement devices contained conservatisms. In reviewing power history back to the end of refueling outage 21, the inspectors noted a few instances, including the excursion on January 18 and 19, where the heat balance steady state exceeded licensed power level. In these cases, power returned to, or less than, the licensed power in a few minutes. Therefore, the inspectors concluded that power returned promptly to the licensed power level and the excursion was not more than minor issue.

This review constituted one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

No findings were identified.

4OA3 Followup of Events and Notices of Enforcement Discretion

.1 Plant Response to a Fault in F Bus Supply Breaker

a. Inspection Scope

On January 8, the power supply breaker to the F bus from the 1-1 start up power transformer failed catastrophically due to water intrusion. The licensee declared an Unusual Event because observers of the failure reported an explosion within the protected area. The loss of the F bus resulted in the loss of one cooling tower and necessitated a downpower to about 55 percent power. The transient also resulted in a loss of the rear bus in the switchyard. The rear bus provides one of the qualified offsite circuits; therefore, the licensee entered Limiting Condition for Operation (LCO) 3.8.1 condition A. The inspectors observed licensee actions in the control room and troubleshooting efforts to verify that the licensee took actions consistent with the plants license. In addition, the inspectors reviewed the circumstances surrounding the EAL call. The inspectors indentified one finding related to a failure to maintain the F bus weatherproof.

This event followup review constituted one sample as defined in IP 71153-05.

b. Findings

Introduction:

A self revealed finding with no associated NCV occurred on January 8 when water intrusion into the F bus resulted in a loss of the rear bus, a downpower due to loss of one cooling tower and an entry into an Unusual Event.

Discussion: On January 8, the plant experienced a transient due to the loss of a nonsafety-related bus that provides power to a cooling tower. Water intrusion into a breaker cubicle caused an internal fault that resulted in loss of the rear bus and loss of a cooling tower. The rear bus was part of the switchyard connecting the plant to the electrical grid and formed a portion of a qualified circuit of offsite power. Based on a report from a security officer of an explosion and observation by an Auxiliary Operator, the shift manager declared an unusual event. In response to the loss of the cooling tower, the licensee entered off-normal procedures for loss of condenser vacuum and for rapid power reduction. The licensee reduced power to ~55 percent in response to the loss of the cooling tower and entered LCO 3.8.1 Condition A due to the loss of one offsite circuit.

The licensee conducted an immediate evaluation of the failed breaker and determined the failure most likely occurred due to water intrusion causing internal faults. The licensee isolated the F bus electrically to complete repairs and to further diagnose the cause of the fault. In addition, the licensee cleared snow and ice from the structure housing the F and G busses to ensure that the G bus would not suffer a similar condition.

In the root cause, the licensee concluded the event resulted from lack adequate preventive maintenance on outdoor switchgear. This inadequacy allowed degradation of the weather seals which that allowed water to enter bus work. In 1995, the licensee experienced similar failure of the G bus. As corrective action to preclude recurrence the licensee implemented a preventive maintenance activity to inspect and repair as necessary caulking on the roof of the F and G bus enclosure. The licensee determined that this Preventative Maintenance (PM) had been cancelled in 2002 but could not locate documentation as to why the PM was cancelled.

Analysis:

The inspectors concluded that the failure to maintain the enclosure was a performance deficiency that warranted a significance determination. The inspectors identified that the licensee failed to maintain the structure weather proof as described in design basis document 3.03 and cancelled a PM activity imposed in C-PAL- 95-0423.

The finding was more than minor because it impacted the initiating event cornerstone objective of limiting the likelihood of those events that upset plant stability and is associated with the attribute of equipment performance. Using IMC 0609, Appendix A, the inspectors determined the finding was of very low safety significance because even though the issue impacted the transient initiating event frequency it did not impact the mitigating system availability. There was no cross-cutting aspect because the cancellation of the PM occurred in 2002 and no longer represents current performance.

From 2002 thru January 8, 2011, the licensee failed to meet self-imposed requirements related to maintaining the structure weatherproof and performing inspection and repairs implemented as corrective action for the 1995 G bus failure.

Enforcement:

This finding does not involve enforcement action because no regulatory requirement violation was identified. Because this finding does not involve a violation and has very low safety significance, it is identified as FIN-05000255/2011002-03, Failure to Maintain Switchgear Weatherproof. The licensee entered the finding into the licensees CAP as CR-PLP-2011-104.

.2 Plant Response Due to a Plant Trip

a. Inspection Scope

On January 22, the inspectors responded to the site due to a plant trip. At 1735 the plant tripped on loss of load. Subsequent investigation revealed that a fault on power cable between the 1-3 Station power transformer and G bus caused a current surge which tripped the turbine. After arriving onsite, the inspectors observed operator performance in completing the actions required by EOP-1, Standard Post Trip Actions.

During the transient, a blown fuse prevented use of the steam dumps to the condenser as a means of decay heat removal. The licensee relied on the use of the atmospheric steam dumps until the fuse was replaced. The inspectors will review the cause of the trip and the response of the steam dumps to condenser as subsequent inspection activity.

b. Findings

No findings were identified This event followup review constituted one sample as defined in IP 71153-05.

4OA5 Other Activities

.1 (Closed) Temporary Instruction 2515/179, Verification of Licensee Responses to NRC

Requirement for Inventories of Materials Tracked in the National Source Tracking System Pursuant to Title 10, Code of Federal Regulations, Part 20.2207 (10 CFR 20.2207)

a. Inspection Scope

The inspectors confirmed that the licensee reported the initial inventories of sealed sources pursuant to 10 CFR 20.2207 and verified that the National Source Tracking System database correctly reflected the Category 1 and 2 sealed sources in custody of the licensee. Inspectors interviewed personnel and performed the following:

• reviewed the licensees source inventory;

• verified the presence of any category 1 or 2 sources ;

• reviewed procedures for and evaluated the effectiveness of storage and handling of sources;

• reviewed documents involving transactions of sources; and

• assessed the adequacy of licensee maintenance, posting, and labeling of nationally tracked sources.

b. Findings

No findings were identified.

.2 (Closed) Unresolved Item 05000255/2010004-07, Description of Liquid Waste Incidents

a. Inspection Scope

The inspectors reviewed the licensees evaluation of the liquid radioactive waste incident description in the UFSAR. The inspectors concluded that the license requires that tanks T-90 and 91 maintain concentrations of tritium less than 1000 times the limits in 10 CFR 20 Appendix B. These requirements are discussed in Section 11.2 of the UFSAR. However, the licensee did not translate these requirements into procedures.

In addition, the inspectors determined that section 14.20 required clarification to describe the liquid waste incident.

b. Findings

Introduction:

The inspectors identified a finding and associated NCV of TS 5.4 for failure to establish and implement procedures recommended by Regulatory guide 1.33.

Specifically, the licensee failed to establish procedures for liquid radioactive waste and emergency procedures for abnormal releases of radioactivity.

Discussion: The inspectors reviewed several licensing documents related to tanks T-90 and T-91. These permanently installed tanks were located outside of the containment and auxiliary buildings near the shore of Lake Michigan. These tanks contain water contaminated with tritium. Although the water is processed to remove radioactive constituents, it still contains tritium because tritium cannot be chemically separated from water. These tanks have exhibited leaks which resulted in the release of tritium-contaminated water to the environment. No instances were identified where detectable tritium was found in drinking water. In addition, there were no indentified releases where 10 CFR part 20 limits for were exceeded.

The inspectors reviewed UFSAR Section 14.20 which discussed the radiological consequences of an accidental liquid waste release; however, the discussion was limited to releases to the circulating canal of a gaseous release from the volume control tank.

Section 14.20 stated that administrative controls, including controls on tanks T-90 and 91, provided assurance that no releases to the environs would exceed 10 CFR 20 limits.

However, the licensee did not implement controls on the content of tanks T-90 and 91 and; therefore, did not limit or evaluate the acceptable concentration of tritium in these tanks.

The inspectors also reviewed the licensee procedure for radioactive spills. The licensee does not have a site specific procedure to address incidents from tank T-90 or T-91.

However, the licensee does have a fleet procedure, EN-RP-113, Response to Contaminated Spills/Leaks which provides guidance for response to spills. The inspectors asked operators about the procedure and concluded that operators, in general, were not aware of the procedure. In addition, the procedure was not part of the operator training program. Further, the procedure was mainly administrative in nature and outlines documentation, notification, and reporting protocols after a leak or spill was identified. These protocols direct notification of local government officials within one business day after a spill was identified. Since a rupture of either tank could potentially impact offsite locations, the inspectors concluded that the delay of 1 business day may not suffice to ensure the assumptions made in the ODCM are valid.

The inspectors concluded that procedures had not been established and implemented as required to:

1. limit the concentration of radioactive materials in T-90/T-91; 2. mitigate the amount of radioactive material leaked to the unrestricted area; and 3. minimize access to members of the public in an affected unrestricted area.

Analysis:

The inspectors concluded that the failure to maintain procedures as required by TS 5.4 was a performance deficiency that warranted a significance determination.

The inspectors compared the issue to the examples in Appendix E, however, none applied. The inspectors determined the finding was more than minor because it impacted the public radiation safety cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation, in that, the licensee failed to meet the program and process attribute of procedures. The site previously identified leakage from the tank and ancillary piping into environment.

Although known leak sources have been repaired, some tritium remains in on-site sample well samples. No tritium has been detected in areas accessible to the public or in drinking water sources. Using the radioactive material control program branch of the public radiation safety SDP, IMC 0609, Appendix D, the inspectors concluded that since the finding resulted in less than

.005 rem exposure to members of the public, the finding

was of very low safety significance (Green). There was no cross-cutting aspect in that the procedures and UFSAR content have been in place for several years and do not reflect current plant performance.

Enforcement:

Technical Specification 5.4 requires, in part, that the licensee implement and maintain procedures recommended in regulatory Guide 1.33. Regulatory Guide 1.33 recommends procedures for the liquid radioactive waste system. Contrary to this requirement, on December 12, 2010, the licensee did not maintain procedures for sampling effluents transferred to or stored in tanks T-90 and 91 for tritium to ensure they met assumptions embedded in the accident analysis. The analysis assumed a factor of 1000 dilution. The licensee has entered the finding into their CAP as CR-PLP-2011-4166. Because the violation is of very low safety significance and was entered in to the licensees CAP, this violation is being treated as a NCV consistent with Section 2.3.2 of the NRC enforcement Policy. NCV 05000255/2011002-04, Inadequate Procedural Controls for Liquid Radioactive Waste.

4OA6 Management Meetings

.1 Exit Meeting Summary

On April 13, 2011, the inspectors presented the inspection results to T. Kirwin and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

.2 Interim Exit Meetings

Interim exits were conducted for:

• The results of Occupational Dose Assessment inspection with the Site Vice President, T. Kirwin, on February 18, 2011.

The inspectors confirmed that none of the potential report input discussed was considered proprietary. Proprietary material received during the inspection was returned to the licensee.

4OA7 Licensee Identified Violations

The following violation of very low significance (Green) or Severity Level IV was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV.

In July of 2010, the licensee identified that the neutron fluence limit for the reactor vessel would be reached as soon as May 2011. Previously, the licensee believed that the limit would be reached in 2014. The licensee determined that the change resulted from the incorrect determination of fluence limits in the governing document for reactor vessel monitoring. The inspectors concluded that the failure to include correct limits for vessel fluence in governing procedures was a violation of 10 CFR 50 Appendix B, Criterion V.

Criterion V requires, in part, that activities affecting quality be prescribed by procedures appropriate to the circumstances. Contrary to this requirement, the site procedure was not appropriate to the circumstances in that it contained incorrect fluence limits. Based on corrected fluence limits, the licensee determined that the curves were valid until May 2011. Further analysis by the licensee provided sufficient basis to extend applicability of the pressure-temperature curves until March 2012. The licensee has submitted a license amendment that uses recent changes to American Society of Mechanical Engineering code to revise the pressure temperature curves to extend the time further.

The inspectors determined that the failure to properly monitor vessel fluence was a performance deficiency that warranted a significance determination. The inspectors concluded that the issue was more than minor because, if left uncorrected, the licensee would have operated without a valid analysis to support the pressure/temperature limits.

Since the licensee identified and corrected the analysis before invalidating the existing analysis, the inspectors concluded the finding was not of more than very low safety significance. The licensee entered this condition into their CAP as CR-PLP-2010-3252.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

M. Ginzel, Health Physicist

C. Sherman, Radiation Protection Manager

Nuclear Regulatory Commission

J. Giessner, Chief, Reactor Projects Branch 4

B. Dickson, Chief, Plant Support Team

LIST OF ITEMS

Opened

05000255/2011002-01 NCV Failure to Account for Potential Age-Related Degradation in EDG Governors (1R15)

05000255/2011002-02 NCV Violation of Fatigue Rule Requirements (4OA2)

05000255/2011002-03 FIN Failure to Maintain Switchgear Weather Proof (4OA3)

05000255/2011002-04 NCV Inadequate Procedural Controls for Liquid Radioactive Waste (4OA5)

Closed

05000255/2011002-01 NCV Failure to Account for Potential Age-Related Degradation in EDG Governors (1R15)

05000255/2011002-02 NCV Violation of Fatigue Rule Requirements (4OA2)

05000255/2011002-03 FIN Failure to Maintain Switchgear Weather Proof (4OA3)

05000255/2011002-04 NCV Inadequate Procedural Controls for Liquid Radioactive Waste (4OA5)

05000255/2010004-07 URI Description of Liquid Waste Incidents (4OA5)

Discussed

None Attachment