IR 05000285/2010004

IR 05000285-10-004, on 07/01/2010 - 09/30/2010, Fort Calhoun, Radiological Hazard Assessment & Exposure Control; Occupational ALARA Planning & Controls; Operability Evaluations; Evaluations of Changes, Tests or Experiments & Permanent Modif

ML103130148
Person / Time
Site:	Fort Calhoun
Issue date:	11/08/2010
From:	Clark J NRC/RGN-IV/DRP/RPB-E
To:	Bannister D Omaha Public Power District
References
IR-10-004
Download: ML103130148 (88)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION R E GI ON I V 612 EAST LAMAR BLVD, SUITE 400 ARLINGTON, TEXAS 76011-4125

November 8, 2010

David J. Bannister, Vice President and Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station FC-2-4 P.O. Box 550 Fort Calhoun, NE 68023-0550

Subject: FORT CALHOUN - NRC INTEGRATED INSPECTION REPORT 05000285/2010004

Dear Mr. Bannister:

On September 30, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Fort Calhoun Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on October 18, 2010, with Mr. J. Reinhart, Site Vice President, and other members of your staff.

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

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

This report documents three NRC-identified findings of very low safety significance (Green), four self-revealing findings of very low safety significance (Green) and three NRC-identified Severity Level IV violations. All 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 listed in this report. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as noncited violations, consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest the violations or the significance of the noncited violations, 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, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspectors at the Fort Calhoun Station facility. In addition, if you disagree with the crosscutting 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 IV, and the NRC Resident Inspectors at the Fort Calhoun Station.

OMAHA PUBLIC POWER DISTRICT

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

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

Sincerely,

/RA/

Mr. Jeffery A. Clark, P.E.

Chief, Project Branch E Division of Reactor Projects

Docket: 50-285 License: DPR-40

Enclosure:

NRC Inspection Report 05000285/2010004 w/Attachment: Supplemental Information

REGION IV==

Docket:

05000285 License:

DPR-40 Report:

05000285/2010004 Licensee:

Omaha Public Power District Facility:

Fort Calhoun Station Location:

9610 Power Lane Blair, NE 68008 Dates:

July 1 through September 30, 2010 Inspectors:

J. Kirkland, Senior Resident Inspector J. Wingebach, Resident Inspector W. Schaup, Project Engineer R. Hickok, Reactor Technology Instructor D. Stearns, Health Physicist N. Greene, Health Physicist G. George, Reactor Inspector, Engineering Branch 1 I. Anchondo, Reactor Inspector, Plant Support Branch 2 T. Buchanan, Reactor Inspector, Plant Support Branch 2 L. Ricketson, Senior Health Physicist D. Stearns, Health Physicist D. Graves, Health Physicist

Approved By:

Jeffrey Clark, P.E., Chief, Project Branch E Division of Reactor Projects

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Enclosure

SUMMARY OF FINDINGS

IR 05000285/2010004; 07/01/2010 - 09/30/2010; Fort Calhoun Station; Radiological Hazard

Assessment and Exposure Control; Occupational ALARA Planning and Controls; Operability Evaluations; Evaluations of Changes, Tests or Experiments and Permanent Modifications;

Identification and Resolution of Problems; and Event Follow-up.

The report covered a 3-month period of inspection by resident inspectors and three announced baseline inspections by region-based inspectors. Seven Green noncited violations of significance and three Severity Level IV violations were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination Process; the crosscutting aspect was determined using Inspection Manual Chapter 0310, Components within the Crosscutting Areas. Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

Cornerstone: Initiating Events

Green.

The inspectors reviewed a self-revealing Green noncited violation of Fort Calhoun Station Technical Specification 5.8.1, for the licensees failure to provide an adequate maintenance procedure for fire protection system flushing.

Specifically, while performing OP-PM-FP-1000 on August 19, 2010, water backed up the VA-87 drain line and spilled onto the east switchgear room floor, into Room 19 below, as well as pooling on top of and inside of cable trays. The licensee has entered this issue into their corrective action program as Condition Report 2010-4423.

The inadequate maintenance procedure is a performance deficiency. This finding is more than minor because if left uncorrected the performance deficiency could have the potential to lead to a more significant safety concern. Specifically the use of OP-PM-FP-1000 allows the potential wetting of safety related equipment in the east switchgear room and Room 19. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609, Appendix A, to determine its significance.

Using Attachment 4 of that appendix, the inspectors determined that the finding has very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. Through conversations with the fire protection system engineer and other licensee members and the fact that similar issues have occurred in the past, the inspectors determined that the primary cause of this finding was the failure to adequately assess the significance of previous condition reports which would have required them to perform a more thorough cause evaluation. Therefore, this finding has a crosscutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems, such that, the resolutions address causes and extent of conditions, as necessary

P.1(c)(Section 4OA2).

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green noncited violation of 10 CFR Part 50,

Appendix B, Criteria III due to the failure of the licensee to perform suitable testing to determine the adequacy of the design of equipment related to transferring diesel fuel from one storage tank to another. Specifically, the inspectors questioned whether fuel oil transfer pump FO-37 or a portable hand pump to be used in the event that FO-37 was unavailable to transfer fuel from storage tank FO-10 to FO-1 would be able to perform the design function. No calculations or previous testing documentation could be provided and when tested to demonstrate that the portable hand pump could perform the intended design function, the portable hand pump failed. Subsequently, the licensee evaluated that fuel oil transfer pump FO-37 is adequately designed to transfer fuel oil from FO-10 to FO-1. The licensee entered this issue into the corrective action program as Condition Reports 2010-3123, 2010-3921, and 2010-4315.

The inspectors determined that the licensees failure to provide calculations or testing documentation that fuel oil transfer pump FO-37 or the designated portable hand pump could perform the intended design function was a performance deficiency. This finding is greater than minor because it affected the Mitigating System Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the equipment performance attribute to maintain availability and reliability of the diesel generators. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has a very low safety significance (Green) because it was not a design or qualification deficiency, does not represent an actual loss of safety function nor did it screen as potentially risk significant for external events. Since the finding is not indicative of current licensee performance, there is no crosscutting aspect assigned to this finding (Section 1R15).

Green.

The inspectors identified a Green noncited violation of Technical Specification 5.8.1(a) for inadequate procedures associated with 4160 V and 480 V safety-related breaker maintenance procedures. The inspectors determined that maintenance procedures used to ensure that 4160 V and 480 V safety-related breakers were being maintained and overhauled in a timely manner were inadequate. The licensee did not have an engineering analysis or technical basis to justify the deviation from vendor and/or Electric Power Research Institute guidance. The inspectors determined that this issue affected the procedure quality attribute for maintenance procedures of the Mitigating System

Cornerstone of reactor safety. Specifically, the issue was more than minor because the failure to incorporate the vendor required maintenance and frequency or fully incorporate Electric Power Research Institute maintenance recommendations for extending the service interval into maintenance procedures for safety related breakers. If left uncorrected, this failure affected the availability, reliability, and capability of mitigating systems that respond to initiating events to prevent undesirable consequences because the reliability of safety-related breakers refurbished using the deficient procedures cannot be predicted. This issue was entered into the licensees corrective action program as Condition Report 2009-2306.

Using the Significance Determination Process, Phase 1 Screening Worksheet, for the Initiating Events, Mitigating Systems, and Barriers Cornerstones the finding was potentially risk significant for multiple systems. Because the probability of multiple system effects is not effectively addressed by a Phase 2 analysis, a Phase 3 analysis was performed. The analyst determined that while the licensee failed to perform adequate maintenance on the breakers, the actual failure rate of the breakers was no greater than the theoretical design failure rate.

The finding was determined to be of very low safety significance because the deficiency did not result in any loss of function. The finding was not risk significant due to a seismic, flooding, or severe weather-initiating event and because other plant-specific analyses that identify core damage scenarios of concern were not impacted. This finding has a crosscutting aspect in the area of problem identification and resolution because the licensee did not effectively incorporate pertinent industry operating experience into the preventive maintenance programs for the 4160 V and 480 V safety-related and risk significant non-safety-related circuit breakers P.2(b)(Section 4OA2)

Green.

A self-revealing Green noncited violation of Fort Calhoun Station Technical Specification 5.8.1 occurred for an inadequate procedure for verifying the connection between cable lugs and cables. This inadequacy resulted in the loss of Motor Control Center MCC-3A1 and a subsequent plant shutdown. The licensee repaired the affected equipment and entered this issue into the corrective action program as Condition Report 2010-4423.

The inspectors determined that the licensees inadequate maintenance procedure was a performance deficiency. This finding was greater than minor because it was similar to a non-minor example 4.b in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that a procedural error caused a reactor trip or other transient. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has very low safety significance because all of the items in Table 4a, of the Mitigating Systems Cornerstone checklist, were answered in the negative. Since the finding is not indicative of current licensee performance, there is no crosscutting aspect assigned to this finding (Section 4OA3).

Cornerstone: Barrier Integrity

Green.

The inspectors identified a noncited violation of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, the licensee failed to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, since January 2009, the licensee failed to correctly translate results of Calculation FC 05561, CCW Relief Valve Setpoints, into calibration procedures used to calibrate pressure control switches PCS-412 and PCS-413. The licensee has entered this violation into their corrective action program as Condition Report 2010-3658.

The inspectors determined that the failure to correctly translate the results of the setpoint calculation into calibration procedures and instructions as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control is a performance deficiency. The finding was more than minor because it adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Additionally, the finding was more than minor because the finding resulted in a condition where there was a reasonable doubt on the operability of the component cooling water system containment isolation valves.

Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the finding was determined to have very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building. This finding has a crosscutting aspect in the area of human performance work practice because the licensee failed to define and effectively communicate expectations regarding procedural compliance and personnel following procedures. Specifically, in January 2009, the licensee failed to effectively communicate expectations regarding personnel following procedures to implement calculation changes H.4(b)(Section 1R17).

Cornerstone: Occupational Radiation Safety

Green.

The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.8.1, for failure to follow radiation work permit requirements. On November 13, 2009, two individuals became contaminated while cleaning the gasket seating surface on the endbell of the letdown heat exchanger because they did not use face shields as required by the radiation work permit. The licensee immediately restricted the two individuals from entry into the radiologically controlled area, conducted a coaching session with the individuals involved and placed this issue into the corrective action program as Condition Report 2009-5688.

The failure to follow the instructions listed on a radiation work permit was a performance deficiency. The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to follow radiation work permit instructions increased personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work practices, human error prevention techniques, because the individuals failed to use self and peer checking to ensure they were signed onto the appropriate task for the work to be performed H.4(a)(Section 2RS01).

Green.

The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.8.1, for failure to appropriately control radiation exposures due to improperly planned maintenance activities associated with Work Package 09-AP-20. The maintenance work involved valve modifications and boric acid system cleanups. These activities resulted in exceeding the original dose estimate by more than 50 percent. The licensee entered this issue into the corrective action program as Condition Reports 2009-6171, 2009-6264 and 2010-1696.

The failure to properly plan maintenance activities to minimize personnel radiation dose is a performance deficiency. This finding is greater than minor because it affected the Occupational Radiation Safety Cornerstone attribute of program and process in that ALARA planning or radiological controls did not prevent unplanned, unintended dose for a work activity. This caused increased collective radiation dose for the job activity to exceed the planned dose of approximately 14 rem by more than 50 percent. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this finding to be of very low safety significance because the finding involved ALARA planning and controls and the licensees latest rolling 3-year average does not exceed 135 person-rem. This finding had an associated human performance crosscutting aspect in the work practices component because the licensee did not ensure supervisory and management oversight of work activities, including the contractor, to maintain doses ALARA H.4(c)(Section 2RS02).

Cornerstone: Miscellaneous

• Severity Level IV. The inspectors identified a Severity Level IV noncited violation for the failure to submit a licensee event report within 60 days as required by 10 CFR 50.73. Specifically, the diesel fuel oil storage system was inoperable for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from January 6, 2010, until January 7, 2010. On January 6, 2010, fuel oil transfer pump FO-37 was inoperable due to a fire main rupture submerging the pump for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With no other means to transfer fuel from storage tank FO-10 to FO-1, the fuel oil storage system was inoperable, and the fuel volume in FO-10 was unavailable. This was reportable condition required by 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by technical specifications. The licensee performed a reportability evaluation, and the violation was entered into the corrective action program as Condition Report 2010-3865.

The inspectors determined that the licensees failure to submit a licensee event report was a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC's regulatory ability was potentially affected. Specifically, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore more than minor. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for traditional enforcement only in accordance with the NRC Enforcement Policy. This is a Severity Level IV violation as defined in Section 2.2.1.c of the NRC Enforcement Policy (Section 1R15).

• Severity Level IV. The inspectors identified a Severity Level IV, noncited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, paragraph (e)which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Contrary to the above, the licensee failed to update periodically the Updated Safety Analysis Report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed.

Specifically, since December 2006, the licensee stored a significant source of radioactivity in the original steam generator storage facility but failed to describe the source, volume, and storage of radioactive equipment in the Updated Safety Analysis Report. The licensee has entered this violation into their corrective action program as Condition Report 2010-3636.

The inspectors determined that the failure to update the Updated Safety Analysis Report as required by 10 CFR 50.71(e), Maintenance of Records, Making of Reports was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding was more than minor because it had a material impact on licensed activities in that a radioactive solid waste storage facility was relocated from the plant radiological controlled area to the owner controlled area without being described in the Updated Safety Analysis Report. The finding was characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy (Section 1R17).

• Severity Level IV. The inspectors identified a Severity Level IV violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the facility as described in the Updated Safety Analysis Report. On April 9, 2010, the licensee changed the facility as described in the Updated Safety Analysis Report to install a cable splice in a safety related cable without determining if prior NRC approval was required. The licensee took actions to make the modification temporary until a permanent repair could be made and entered the issue into the corrective action program as Condition Report 2010-4466.

Fort Calhoun Station utilizes NEI 96-07 as their process to meet 10 CFR 50.59 requirements. Their failure to perform a 10 CFR 50.59 evaluation, in accordance with NEI 96 07, prior to changing the facility as described in the Updated Safety Analysis Report is a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC's regulatory ability was potentially affected. Specifically, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore more than minor. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for Traditional Enforcement only in accordance with the NRC Enforcement Policy. The inspectors concluded that the 10 CFR 50.59 evaluation would have likely identified that prior NRC approval would have been required, unless the change to the facility was for a short duration of time. This was due to the introduction of additional potential failure mechanisms of the splices that are age-dependent.

Since the licensee subsequently classified the cable splice as a temporary modification, and scheduled to be removed during the next refueling outage, the aging mechanisms would no longer be applicable. Therefore, this is a Severity Level IV violation as defined in Section 2.2.1.c of the NRC Enforcement Policy (Section 1R20).

Licensee-Identified Violations

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 action tracking numbers (condition report numbers) are listed in Section 4OA7.

REPORT DETAILS

Summary of Plant Status

The plant operated at approximately 100 percent power during the entire inspection period.

REACTOR SAFETY

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

1R01 Adverse Weather Protection

.1 Summer Readiness for Offsite and Alternate-ac Power

a. Inspection Scope

The inspectors performed a review of preparations for summer weather for selected systems, including conditions that could lead to loss-of-offsite power and conditions that could result from high temperatures. The inspectors reviewed the procedures affecting these areas and the communications protocols between the transmission system operator and the plant to verify that the appropriate information was being exchanged when issues arose that could affect the offsite power system. Examples of aspects considered in the inspectors review included:

• The coordination between the transmission system operator and the plants operations personnel during off-normal or emergency events

• The explanations for the events

• The estimates of when the offsite power system would be returned to a normal state

• The notifications from the transmission system operator to the plant when the offsite power system was returned to normal

During the inspection, the inspectors focused on plant-specific design features and the procedures used by plant personnel to mitigate or respond to adverse weather conditions. Additionally, the inspectors reviewed the Updated Safety Analysis Report and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant-specific procedures. Specific documents reviewed during this inspection are listed in the attachment. The inspectors also reviewed corrective action program items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures. The inspectors reviews focused specifically on the following plant systems:

• August 20, 2010, 345 kV switchyard, 13.8 kV transformers T1-A3 and T1-A4

This activity constitutes completion of one

(1) sample for readiness for summer weather affect on offsite and alternate-ac power as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings were identified.

1R04 Equipment Alignments

.1 Partial Walkdown

a. Inspection Scope

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

• August 26, 2010, Portions of the bearing cooling water system while the diesel-driven auxiliary feedwater pump, FW-54, is out of service

• August 30, 2010, Diesel generator 1 starting air system while diesel generator 2 is inoperable

• September 2, 2010, Portions of the high pressure safety injection system while high pressure safety injection pump SI-2B inoperable

• September 22, 2010, Diesel generator 2 lube oil system while diesel generator 1 is inoperable

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 affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Safety Analysis Report, technical specification requirements, administrative technical specifications, 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 inspected 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 corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four

(4) partial system walkdown samples as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings were identified.

.2 Complete Walkdown

a. Inspection Scope

On September 21, 2010, the inspectors performed a complete system alignment inspection of the containment spray to verify the functional capability of the system. The inspectors selected this system because it was considered both safety significant and risk significant in the licensees probabilistic risk assessment. The inspectors inspected the system to review mechanical and electrical equipment line ups, electrical power availability, system pressure and temperature indications, as appropriate, component labeling, component lubrication, component and equipment cooling, hangers and supports, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. The inspectors reviewed a sample of past and outstanding work orders to determine whether any deficiencies significantly affected the systems function. In addition, the inspectors reviewed the corrective action program database to ensure that system equipment-alignment problems were being identified and appropriately resolved. Specific documents reviewed during this inspection are listed in the attachment.

This activity constitutes completion of one

(1) sample of a complete system walkdown as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

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

• July 13, 2010, Fire Area 1, Shutdown Heat Exchanger Area I, Room 21

• July 13, 2010, Fire Area 6.5, Safety Injection and Containment Spray Pump Area I, Room 15

• July 13, 2010, Fire Area 6.6, Shutdown Heat Exchanger Area II, Room 14

• August 18, 2010, Fire Area 42, Control Room Complex Area

The inspectors reviewed areas to assess if licensee personnel 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 had 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 affect equipment that 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 corrective action program.

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

These activities constitute completion of four

(4) quarterly fire-protection inspection samples as defined in Inspection Procedure 71111.05-05.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

a. Inspection Scope

The inspectors reviewed the Updated Safety Analysis Report, the flooding analysis, and plant procedures to assess susceptibilities involving internal flooding; reviewed the corrective action program to determine if licensee personnel identified and corrected flooding problems; inspected underground bunkers/manholes to verify the adequacy of sump pumps, level alarm circuits, cable splices subject to submergence, and drainage for bunkers/manholes; and verified that operator actions for coping with flooding can reasonably achieve the desired outcomes. The inspectors also inspected the areas listed below to verify the adequacy of equipment seals located below the flood line, floor and wall penetration seals, watertight door seals, common drain lines and sumps, sump pumps, level alarms, and control circuits, and temporary or removable flood barriers.

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

• August 16, 2010, Flooding of onsite manholes

This activity constitutes completion of one

(1) bunker/manhole sample as defined in Inspection Procedure 71111.06-05.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

Quarterly Review

a. Inspection Scope

On July 19, 2010, the inspectors observed a crew of licensed operators in the plants simulator 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

• Crews ability to take timely actions in the conservative direction

• Crews prioritization, interpretation, and verification of annunciator alarms

• Crews correct use and implementation of abnormal and emergency procedures

• Control board manipulations

• Supervisors oversight and direction

• Crews ability to identify and implement appropriate technical specification actions and emergency plan actions and notifications

The inspectors compared the crews performance in these areas to pre-established operator action expectations and successful critical task completion requirements.

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

These activities constitute completion of one

(1) quarterly licensed-operator requalification program sample as defined in Inspection Procedure 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:

• August 11, 2010, Heater drain tank level control valve, LCV-1199, exceeding its performance criteria

• September 14, 2010, Corrective actions associated with breaker 1A31

The inspectors reviewed events such as, where ineffective equipment maintenance has 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)

• Characterizing system reliability issues for performance

• Charging unavailability for performance

• Trending key parameters for condition monitoring

• Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or (a)(2)

• Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as requiring the establishment of appropriate and adequate goals and corrective actions for systems classified as not having adequate performance, as described in 10 CFR 50.65(a)(1)

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified that maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two

(2) quarterly maintenance effectiveness samples as defined in Inspection Procedure 71111.12-05.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed licensee personnel'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:

• August 10, 2010, Maintenance on 345 kV to 161 kV transformer T3 and associated switchyard work

• August 25, 2010, Yellow risk condition associated with low pressure safety injection pump, SI-1A, containment spray pump, SI-3A, and diesel generator 1, all out of service

• August 30, 2010, Emergent work to troubleshoot diesel generator 2 jacket water temperature switch

• September 7, 2010, Yellow risk associated with the removal of potable water to all instrument air compressors.

The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. 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 the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four

(4) maintenance risk assessments and emergent work control inspection samples as defined in Inspection Procedure 71111.13-05.

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• July 27, 2010, Operability of control room air conditioning unit, VA-46B, following failure of fan pressure switch 4511

• July 29, 2010, Operability of Room 81 due to internal flooding

• August 9, 2010, Operability of control room air conditioning unit, VA-46B, compressor fan due to faulty pressure switch

• August 31, 2010, Operability of the manual fuel oil transfer pump dedicated to transfer fuel from storage tank FO-10 to FO-1

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 technical specification 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 technical specifications and Updated Safety Analysis Report to the licensee personnels 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 also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four

(4) operability evaluations inspection samples as defined in Inspection Procedure 71111.15-04

b. Findings

(1)

Introduction.

The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III due to the failure of the licensee to perform suitable testing to determine the adequacy of the design of equipment related to transferring diesel fuel from one storage tank to another. Specifically, the inspectors questioned whether fuel oil transfer pump FO-37 or a portable hand pump to be used in the event that FO-37 was unavailable to transfer fuel from storage tank FO-10 to FO-1 would be able to perform the design function.

No calculations or previous testing documentation could be provided and when

tested to demonstrate that the portable hand pump could perform the intended design function the portable hand pump failed.

Description.

Fort Calhoun Station Technical Specification 2.7.1 describes minimum requirements for electrical systems, including the volume of available fuel required for the diesel generators. Prior to 1994, the minimum requirement was a full engine base day tank and a minimum of 16,000 gallons of fuel in the underground storage tank, FO-1.

In September 1988, during a design basis reconstitution effort, the licensee identified that the fuel oil inventory currently maintained in the fuel oil storage tank FO-1 was no longer sufficient for 7 days of continuous diesel generator operation following a loss of coolant accident. This was based upon a revised fuel oil calculation and additional loads, resulting from meeting post Three Mile Island requirements, being added to the emergency diesel generators. The licensee determined a minimum inventory of 24,520 gallons of fuel oil would be required in order to ensure 7 days of continuous diesel generator operation at the loads necessary to support and maintain safe reactor shutdown during the most limiting accident.

On September 17, 1993, the licensee submitted an application for amendment of its operating license to comply with the new minimum inventory requirements. In order to comply, the technical specification would require a minimum of 16,000 gallons of fuel in storage tank FO-1, and a minimum of 8,000 gallons of fuel in the auxiliary boiler fuel oil storage tank, FO-10. The amendment request described two methods of transferring fuel from FO-10 to FO-1. The first described use of the diesel-driven auxiliary feedwater pump fuel oil transfer pump, FO-37, through manually installed hoses. The second described using a portable pump through manually installed hoses. The amendment request stated that the methods of transfer of the fuel oil from storage tank FO-10 to FO-1 had been established and procedures had been developed so that the transfer could be made in a timely manner without adversely impacting diesel generator operation. The amendment request was approved on March 29, 1994. A subsequent amendment to the technical specification increased the required fuel available in FO-10 to 10,000 gallons.

After a walkdown and review of emergency diesel generator 2 on June 9, 2010, the inspectors requested to see the calculations or testing documentation used to determine the adequacy of the design of fuel oil transfer pump FO-37 and the portable hand pump designated as the backup for FO-37. The licensee determined that no calculations were performed and that no testing was performed for either pump as required per 10 CFR Part 50, Appendix B, Criteria III. Additionally, when tested to demonstrate that the portable hand pump could perform the intended design function the portable hand pump failed.

Subsequently, the licensee has provided documentation that fuel oil transfer pump FO-37 is adequately designed to transfer fuel oil from storage tank FO-10 to FO-1

Analysis.

The inspectors determined that the licensees failure to provide calculations or testing documentation that fuel oil transfer pump FO-37 or the designated portable hand pump could perform the intended design function was a performance deficiency. This finding is greater than minor because it affected the Mitigating System Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the equipment performance attribute to maintain availability and reliability of the diesel generators. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has a very low safety significance (Green) because it was not a design or qualification deficiency, does not represent an actual loss of safety function nor did it screen as potentially risk significant for external events. Since the finding is not indicative of current licensee performance, there is no crosscutting aspect assigned to this finding.

Enforcement.

Title 10 CFR Part 50, Appendix B, Criteria III, requires in part, that the licensee provide for verifying or checking the adequacy of design, such as, by the performance of design reviews, by the use of alternate or simplified calculation methods, or by the performance of a suitable testing program.

Contrary to the above, no calculations or testing documentation exists to determine the adequacy of the design of fuel oil transfer pump FO-37 or the designated portable hand pump for transferring fuel oil from storage tank FO-10 to FO-1. This performance deficiency has existed since March 29, 1994.

Because the violation was of very low safety significance, there were no actual safety consequences and was entered into the licensee's corrective action program as Condition Reports 2010-3123, 2010-3921 and 2010-4315. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000285/2010004-01, Inadequate Documentation of the Adequacy of Design for the Pumps that Transfer Fuel Oil from Storage Tank FO-10 to FO-1.

(2)

Introduction.

The inspectors identified a Severity Level IV noncited violation for the failure to submit a licensee event report within 60 days as required by 10 CFR 50.73. Specifically, the diesel fuel oil storage system was inoperable for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from January 6, 2010, until January 7, 2010. On January 6, 2010, fuel oil transfer pump FO-37 was inoperable due to a fire main rupture submerging the pump for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With no other mean s to transfer fuel from storage tank FO-10 to FO-1, the fuel oil storage system was inoperable, and the fuel volume in FO-10 was unavailable. This was a reportable condition required by 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by technical specifications.

Description.

Prior to 1994, the licensee technical specification described the minimum requirements for diesel generators as both diesel generators, with a full engine base day tank and a minimum of 16,000 gallons of fuel in the underground storage tank. To meet post Three Mile Island requirements for increased diesel fuel consumption, the licensee implemented Amendment 162 to their technical specifications. The minimum requirements were split into two separate requirements, one of which was One diesel fuel oil storage system containing a minimum volume of 16,000 gallons of diesel fuel in storage tank FO-1, and an additional 8,000 gallons of diesel fuel in storage tank FO-10. This requirement has since been revised to require 10,000 gallons of diesel fuel in FO-10.

In order to transfer diesel fuel to the diesel generators for use, the fuel is transferred from storage tank FO-1 to the diesel generator day tanks, thus in order to use the fuel in storage tank FO-10, that fuel must be first transferred to FO-1. In the amendment request, the licensee addressed how this would be accomplished. A modification was made to the diesel-driven auxiliary feedwater pump, FW-54, to allow use of its fuel oil transfer pump, FO-37, to transfer fuel from storage tank FO-10 to FO-1 via a dedicated portable hose with sufficient length to connect the two storage tanks. The licensee also indicated that, as a backup to fuel oil transfer pump F0-37, a dedicated and tagged portable pump will be provided and stored in an appropriate area. Transferring the fuel from storage tank FO-10 to FO-1 is described in Procedures EPIP-RR-17A, TSC Administrative Logistics Coordinator Actions, and the EOP/AOP Attachments.

On January 6, 2010, a fire main rupture in the service building submerged fuel oil transfer pump FO-37 and it was declared inoperable, and remained inoperable for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The inspectors questioned the licensee on June 10, 2010, regarding the operability and reportability of the condition where fuel oil transfer pump FO-37 is inoperable, leaving the fuel volume in storage tank FO-10 as unavailable. The licensee responded that with FO-37 inoperable, the portable pump was available for fuel transfer, thus the fuel in storage tank FO-10 was available, the diesel generators and fuel oil storage systems were operable, and there was no reportable condition.

On June 23, 2010, the inspectors further questioned the ability of the portable pump to transfer the fuel from storage tank FO-10 to FO-1. Following a review of the equipment, the licensee determined that the portable pump could not be relied upon to transfer fuel. With no portable pump available, the only method to transfer fuel from storage tank FO-10 to FO-1 would be with fuel oil transfer pump FO-37.

The technical specifications minimum requirement regarding diesel fuel is one diesel fuel oil storage system, with a minimum of 16,000 gallons in storage tank FO-1 and 10,000 gallons in storage tank FO-10. With no means to transfer fuel from FO-10 to FO-1, the diesel fuel oil storage system is inoperable, and the fuel in FO-10 is unavailable. This condition is prohibited by technical specifications

unless the plant is shutdown and less than 300°F in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. On January 6, 2010, the plant remained in Mode 1 at 100 percent power during the entire duration of FO-37, fuel oil transfer pumps, inoperability. Therefore, the condition was prohibited by technical specifications, and was reportable in accordance with 10 CFR 50.73(a)(2)(v) within 60 days.

Analysis.

The inspectors determined that the licensees failure to submit a licensee event report was a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC's regulatory ability was potentially affected. Specifically, the NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore more than minor. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for traditional enforcement only in accordance with the NRC Enforcement Policy. This is a Severity Level IV violation as defined in Section 2.2.1.c of the NRC Enforcement Policy.

Enforcement.

Title 10 CFR 50.73(a)(2)(i)(B) requires, in part, that the licensee shall submit a licensee event report within 60 days of Any operation or condition which was prohibited by the plant's technical specifications. Contrary to the above, the licensee failed to submit a licensee event report within 60 days after discovery of a condition that was prohibited by the licensee technical specifications. The condition prohibited by the technical specifications occurred on January 6, 2010. The inspectors identified the failure to submit a licensee event report on June 10, 2010, and the licensee completed a reportability evaluation on September 9, 2010 and determined that the event was reportable.

There was no actual or potential safety consequences associated with this violation. This is a Severity Level IV noncited violation consistent with Section 2.2.1.c of the NRC Enforcement Policy. Because the violation was not considered to be willful nor repetitive, the licensee took action to perform a reportability evaluation, and the violation was entered into the corrective action program as condition report 2010-3865, this violation is being treated as a Severity Level IV noncited violation, consistent with the NRC Enforcement Policy:

NCV 05000285/2010004-02, Failure to Submit a Required Licensee Event Report

==1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications (71111.17)a.

==

Inspection Scope

The inspectors reviewed the effectiveness of the licensees implementation of evaluations performed in accordance with 10 CFR 50.59, Changes, Tests,

and Experiments, and changes, tests, experiments, or methodology changes that the licensee determined did not require 10 CFR 50.59 evaluations.

The inspectors reviewed 6 evaluations required by 10 CFR 50.59; 15 changes, tests, and experiments that were screened out by licensee personnel; and 8 permanent plant modifications. Documents reviewed are listed in the attachment.

The inspectors verified that, when changes, tests, or experiments were made, evaluations were performed in accordance with 10 CFR 50.59 and licensee personnel had appropriately concluded that the change, test or experiment can be accomplished without obtaining a license amendment. The inspectors also verified that safety issues related to the changes, tests, or experiments were resolved. The inspectors reviewed changes, tests, and experiments that licensee personnel determined did not require evaluations and verified that the licensee personnels conclusions were correct and consistent with 10 CFR 50.59. The inspectors also verified that procedures, design, and licensing basis documentation used to support the changes were accurate after the changes had been made.

In the inspection of modifications, the inspectors verified that supporting design and license basis documentation had been updated accordingly and was still consistent with the new design. The inspectors verified that procedures, training plans, and other design basis features had been adequately accounted for and updated. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of six

(6) samples of evaluations; 15 samples of changes, tests, and experiments that were screened out by licensee personnel; and 8 samples of permanent plant modifications as defined in Inspection Procedure 71111.17-04.

b. Findings

(1) Failure to Update the Updated Safety Analysis Report-Solid Wastes

Introduction.

The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports for failure to update the Updated Safety Analysis Report. Specifically, the licensee built the original steam generator storage facility for solid radioactive waste storage of the two original steam generators, pressurizer, reactor vessel head, and four concrete reactor vessel head missile shield blocks, but failed to update the Updated Safety Analysis Report to include these changes to the facility.

Description.

While reviewing the Fort Calhoun Station Updated Safety Analysis Report, the inspectors identified that the original steam generator storage facility was depicted in an Updated Safety Analysis Report drawing; however, the structure was not described in Chapter 11 or Appendix G of the Updated Safety Analysis Report.

Presently, Chapter 11, Section 11.1.4, Solid Wastes, describes the source, volume, storage, and inspection of non-compactable solid waste, such as equipment parts originating in the control access area of the plant. Criterion 68, Fuel and Waste Storage Radiation Shielding, and Criterion 69, Protection against Radioactivity Release from Spent Fuel and Waste Storage, are provided in Appendix G, Responses to 70 Criteria. Criterion 68 and 69 describe the shielding and containment for radiation protection of solid waste storage as required by 10 CFR Part 20.

The original steam generator storage facility has been in use since 2006 and contains two decommissioned steam generators, a reactor vessel head with control element drive mechanisms, a pressurizer, and four concrete reactor vessel head missile shield blocks. From the licensees estimation, the original steam generator storage facility contains 404 curies, a significant source of radioactivity, not described in the licensees Updated Safety Analysis Report.

The inspectors determined that the two decommissioned steam generators, a reactor vessel head with control element drive mechanisms, a pressurizer, and four concrete reactor vessel head missile shield blocks should be treated as solid radioactive waste, as described in Chapter 11. Additionally, the dedicated storage facility should be described in Criterion 68 and 69 of Appendix G.

However, the licensee failed to include a description of the original steam generator storage facility in Chapter 11 and Appendix G of the Updated Safety Analysis Report.

Analysis.

The inspectors determined that the failure to update the Updated Safety Analysis Report as required by 10 CFR 50.71(e), Maintenance of Records, Making of Reports was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding was more than minor because it had a material impact on licensed activities in that a radioactive solid waste storage facility was relocated from the plant radiological controlled area to the owner controlled area without being described in the Updated Safety Analysis Report. The finding was characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy.

Enforcement.

The inspectors identified a Severity Level IV, noncited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, paragraph (e)which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the Updated Safety Analysis Report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Contrary to the above, the licensee failed to update periodically the Updated Safety Analysis Report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed.

Specifically, since December 2006, the licensee stored a significant source of radioactivity in the original steam generator storage facility but failed to describe

the source, volume, and storage of radioactive equipment in the Updated Safety Analysis Report. The licensee has entered this violation into their corrective action program as Condition Report 2010-3636. This finding was characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy. Because the violation was of Severity Level IV safety significance, this violation is being treated as a noncited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000285/2010004-03, Failure to Update the Updated Safety Analysis Report - Solid Waste.

(2) Failure to Translate Calculation into Calibration Procedure

Introduction.

The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to correctly translate the results of calculation changes into specifications, procedures, and instructions. Specifically, the licensee failed to translate the results of calculation changes into procedures and instructions used to calibrate pressure control switch setpoints in the component cooling water system.

Description.

Pressure control switches PCS-412 and PCS-413 provide component cooling water low pressure input for automatic closure of the inboard and outboard containment isolation valves on component cooling water piping that cools control element drive mechanisms and reactor coolant pump seals.

Each pressure switch is a dual pressure switch that has two pressure setpoints.

The valves automatically close when the plant receives a containment isolation signal coincident with low component cooling water pressure signal from PCS-412 and PCS-413.

In January 2009, the licensee revised Calculation FC 05561, CCW Relief Valve Setpoints, to correct several deficiencies in the previous calculation associated with setpoints for relief valves and pressure control switches in the component cooling water system. One of these revisions changed the recommended setpoint range for pressure control switches PCS-412 and PCS-413. The recommended setpoint range was changed from 27-45 psig to 44-45 psig.

During every refueling outage, the licensee calibrates each pressure control switch using Calibration Procedures IC-CP-01-412/413, Calibration of Pressure Control Switch PCS-412/413, Revision 8. Attachment 9.2 of the procedure states that the desired setpoint is 40 psig with an allowable range between 38 to 44 psig. The inspectors determined that the licensee did not translate results of Calculation FC 05561 into calibration procedures because the desired setpoints in Attachment 9.2 were less than the recommended 44-45 psig range.

In the November 2009, refueling outage, the licensee used Revision 8 of Calibration Procedure IC-CP-01-412/413 to calibrate pressure control switches PCS-412 and PCS-413. The inspectors discovered that the as-left setpoints for the pressure control switches were set less than the recommended 44 psig. The as-left setpoints for the dual pressure switches were set at 43.6/40.2 psig for

PCS-412 and 40.4/40.9 psig for PCS-413. Once this condition was discovered, the licensee entered the condition into Condition Report 2010-3658.

The inspectors determined that the cause of the condition was that the licensee did not effectively implement the expectations set in Calculation Procedure PED-QP-3, Calculation Preparation, Review, and Approval, Revision 24. Section 4.5.6 of PED-QP-3 is used to determine if other documents are affected by the calculation change. This section discusses that the preparer of the calculation shall determine if other documents are impacted by use of PED-QP-3.8, Calculation Affected Documents, Revision 6. The inspectors identified that the preparer marked that the use of PED-QP-3.8 was not applicable to the revision of calculation FC 05561.

Analysis.

The inspectors determined that the failure to correctly translate the results of the setpoint calculation into calibration procedures and instructions as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control is a performance deficiency. The finding was more than minor because it adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Additionally, the finding was more than minor because the finding resulted in a condition where there was reasonable doubt on the operability of the component cooling water system containment isolation valves. Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the finding was determined to have very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building.

This finding has a crosscutting aspect in the area of human performance work practice because the licensee failed to define and effectively communicate expectations regarding procedural compliance and personnel following procedures. Specifically, in January 2009, the licensee failed to effectively communicate expectations regarding personnel following procedures to implement calculation changes H.4(b).

Enforcement.

The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to above, the licensee failed to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, since January 2009, the licensee failed to correctly translate results of Calculation FC 05561, CCW Relief Valve Setpoints, into calibration procedures used to calibrate pressure control switches PCS-412 and PCS-413. The

licensee has entered this violation into their corrective action program as Condition Report 2010-3658. Because this violation is of very low safety significance, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000285/2010004-04, Failure to Translate Calculation into Calibration Procedure.

1R18 Plant Modifications

Temporary Modifications

a. Inspection Scope

To verify that the safety functions of important safety systems were not degraded, the inspectors reviewed the temporary modification identified as replacing temperature input T-123 into the low pressure overpressure protection circuitry.

The inspectors reviewed the temporary modification and the associated safety-evaluation screening against the system design bases documentation, including the Updated Safety Analysis Report and the technical specifications, and verified that the modification did not adversely affect the system operability/availability. The inspectors also verified that the installation and restoration were consistent with the modification documents and that configuration control was adequate. Additionally, the inspectors verified that the temporary modification was identified on control room drawings, appropriate tags were placed on the affected equipment, and licensee personnel evaluated the combined effects on mitigating systems and the integrity of radiological barriers.

These activities constitute completion of one

(1) sample for temporary plant modifications as defined in Inspection Procedure 71111.18-05.

b. Findings

No findings were identified.

1R19 Postmaintenance Testing

a. Inspection Scope

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

• July 14, 2010, Postmaintenance testing of control room air conditioning unit, VA-46A, following compressor failure

• August 12, 2010, Postmaintenance testing of reactor protection system trip unit delta following trip unit replacement

• September 15, 2010, Postmaintenance testing of instrument air compressor, CA-1C, following compressor overhaul

• September 17, 2010, Postmaintenance testing of high pressure safety injection pump SI-2B discharge valve, HCV-2908, and low pressure safety injection pump SI-1B discharge valve HCV-2938 after solenoid replacement

• September 17, 2010, Postmaintenance testing of low pressure safety injection pump SI-1B after system maintenance

The inspectors selected these activities based upon the structure, system, or component's ability to affect 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

The inspectors evaluated the activities against the technical specifications, the Updated Safety Analysis Report, 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 postmaintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of five

(5) postmaintenance testing inspection samples as defined in Inspection Procedure 71111.19-05.

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

.1 (Closed) Unresolved Item 05000285/2010003-06, Failure to Perform a Proper 50.59

Evaluation

a. Inspection Scope

The inspectors identified an unresolved item (URI) during the forced outage which occurred April 8 to April 12, 2010. This URI, (URI 05000285/2010003-06) concerned the 10 CFR 50.59 evaluation of cable splices installed on the power feeder cables for motor

control center MCC-3A1. The inspectors conducted follow-up discussions with personnel from the electrical branch of NRR, as well as inspectors in Region IV Division of Reactor Safety, and determined that a 10 CFR 50.59 evaluation was required. The inspection of the unresolved item did not constitute any additional inspection samples.

Instead, by procedure, they were considered an integral part of the inspections performed during the quarter.

b. Findings

Introduction.

The inspectors identified a Severity Level IV violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the facility as described in the Updated Safety Analysis Report. On April 9, 2010, the licensee changed the facility as described in the Updated Safety Analysis Report to install a cable splice in a safety related cable without determining if prior NRC approval was required.

Description.

On April 9, 2010, the licensee repaired a section of power cable for motor control center MCC-3A1 with cable splices. Approximately 17 feet of 500 MCM cable was removed from each of the three phases for the supply to MCC-3A1 and Burndy compression type butt splices were used to splice new cables to the remaining existing cables.

The inspectors reviewed Section 8.5, Initial Cable Installation Design Criteria of the Updated Safety Analysis Report.

• Updated Safety Analysis Report 8.5 states, in part: The Cable and Conduit Schedule Notes, Figure 8.5-1, provides the standard design criteria for cables and conduits. Deviation from the standard criteria is acceptable provided an analysis has been completed which justified the deviation.

• Updated Safety Analysis Report Figure 8.5-1, Cable and Conduit Schedule Notes, Note 19 states: Splicing in cable trays is not allowed unless specifically called for on drawings. Exceptions to this requirement shall require the written approval of the engineer.

• Updated Safety Analysis Report Figure 8.5-1, Note 26 states: Deviations from the standards stated above is [are] acceptable provided an analysis has been performed to justify the deviation.

• Updated Safety Analysis Report Section 8.5.4.c states: Cable splicing in cable trays is used only for connection of incoming and outgoing cables with containment electrical penetration conductors.

The licensee performed a 50.59 Screen in accordance with the guidance provided in FCSG-23, 10 CFR 50.59 Resource Manual. The guidance adopts NEI 96-07, Revision 1 Guidelines for 10 CFR 50.59 Implementation which includes five screening questions to determine if a complete evaluation of 10 CFR 50.59 is required. The

licensee determined that a cable splice was an equivalent replacement for cable, and thus it screened out in accordance with NEI-96-07 and no evaluation of 10 CFR 50.59 was required. The inspectors determined that a cable splice is not an equivalent replacement, thus a violation of 10 CFR 50.59 occurred for failure to perform an evaluation of the cable splice against the criteria set forth in 10 CFR 50.59 to determine if prior NRC approval was required.

To determine if the splice would require prior NRC approval, the inspectors reviewed Regulatory Guide 1.75, Physical Independence of Electric Systems. Regulatory Guide 1.75 states in part that Section 5.1.1.3 (of IEEE 384-1974) should be supplemented as follows: "(4) Cable splices in raceways should be prohibited."

Updated Safety Analysis Report, Section 8.5.5 states the electrical cables adequately comply with IEEE Standard 383-1974 and meet BTP [Branch Technical Position] 9.5-1.

Further, the Fort Calhoun Station Fire Hazards Analysis states that the requirements of Revision 2 of Regulatory Guide 1.75 (1978) are implemented by design procedures.

The inspectors therefore determined that prior NRC approval was likely required, as cable splices in raceways should be prohibited, in accordance with Regulatory Guide 1.75.

Analysis.

Fort Calhoun Station utilizes NEI 96-07 as their process to meet 10 CFR 50.59 requirements. Their failure to perform a 10 CFR 50.59 evaluation, in accordance with NEI 96 07, prior to changing the facility as described in the Updated Safety Analysis Report is a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC's regulatory ability was potentially affected.

Specifically, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore more than minor. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for Traditional Enforcement only in accordance with the NRC Enforcement Policy. The inspectors concluded that the 10 CFR 50.59 evaluation would have likely identified that prior NRC approval would have been required, unless the change to the facility was for a short duration of time. This was due to the introduction of additional potential failure mechanisms of the splices that are age-dependent. Since the licensee subsequently classified the cable splice as a temporary modification, and scheduled to be removed during the next refueling outage, the aging mechanisms would no longer be applicable.

Therefore, this is a Severity Level IV violation as defined in Section 2.2.1.c of the NRC Enforcement Policy.

Enforcement.

Title 10 CFR 50.59, Changes, Tests and Experiments, states in part that a licensee may make changes in the facility as described in the Updated Safety Analysis Report without obtaining a license amendment if the change does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety

analysis report. Contrary to the above, on April 11, 2010, the licensee changed the facility as described in the Updated Safety Analysis Report to install a cable splice in a safety related cable without determining if prior NRC approval was required. There was no actual or potential safety consequences associated with this violation. This is a Severity Level IV noncited violation consistent with Section 2.2.1.c of the NRC Enforcement Policy. Because the violation was not considered to be willful nor repetitive, the licensee took action to perform a reportability evaluation, and the violation was entered into the corrective action program as Condition Report 2010-4466, this violation is being treated as a Severity Level IV noncited violation, consistent with the NRC Enforcement Policy: NCV 05000285/2010004-05, Failure to Perform a 10 CFR 50.59 Evaluation

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the Updated Safety Analysis Report, procedure requirements, and technical specifications to ensure that the surveillance activities listed below demonstrated that the systems, structures, and/or components tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the significant surveillance test attributes were adequate to address the following:

• Preconditioning

• Evaluation of testing impact on the plant

• Acceptance criteria

• Test equipment

• Procedures

• Jumper/lifted lead controls

• Test data

• Testing frequency and method demonstrated technical specification operability

• Test equipment removal

• Restoration of plant systems

• Fulfillment of ASME Code requirements

• Updating of performance indicator data

• Engineering evaluations, root causes, and bases for returning tested systems, structures, and components not meeting the test acceptance criteria were correct

• Reference setting data

• Annunciators and alarms setpoints

The inspectors also verified that licensee personnel identified and implemented any needed corrective actions associated with the surveillance testing.

• July 18, 2010, OP-PM-AFW-0004, third auxiliary feedwater pump operability verification

• July 20, 2010, RE-ST-RX-0001, determination of total integrated radial peaking factor

• August 25, 2010, OP-ST-RC-3002, reactor coolant system Category B valve exercise test

• August 26, 2010, OP-ST-AFW-3009, auxiliary feedwater pump FW-6, recirculation valve, and check valve tests

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

These activities constitute completion of four

(4) surveillance testing inspection samples as defined in Inspection Procedure 71111.22-05.

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 July 16, 2010, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the Technical Support Center, the Operations Support Center, the Control Room, and in-field observations 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 corrective action program.

As part of the inspection, the inspectors reviewed the drill package and other documents listed in the attachment.

These activities constitute completion of one

(1) sample as defined in Inspection Procedure 71114.06-05.

b. Findings

No findings were identified.

2RS0 1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

This area was inspected to:

(1) review and assess licensees performance in assessing the radiological hazards in the workplace associated with licensed activities and the implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures,

(2) verify the licensee is properly identifying and reporting Occupational Radiation Safety Cornerstone performance indicators, and

(3) identify those performance deficiencies that were reportable as a performance indicator and which may have represented a substantial potential for overexposure of the worker.

The inspectors used the requirements in 10 CFR Part 20, the technical specifications, and the licensees procedures required by technical specifications as criteria for determining compliance. During the inspection, the inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors performed walkdowns of various portions of the plant, performed independent radiation dose rate measurements and reviewed the following items:

• Performance indicator events and associated documentation reported by the licensee in the Occupational Radiation Safety Cornerstone

• The hazard assessment program, including a review of the licenses evaluations of changes in plant operations and radiological surveys to detect dose rates, airborne radioactivity, and surface contamination levels

• Instructions and notices to workers, including labeling or marking containers of radioactive material, radiation work permits, actions for electronic dosimeter alarms, and changes to radiological conditions

• Programs and processes for control of sealed sources and release of potentially contaminated material from the radiologically controlled area, including survey

performance, instrument sensitivity, release criteria, procedural guidance, and sealed source accountability

• Radiological hazards control and work coverage, including the adequacy of surveys, radiation protection job coverage, and contamination controls; the use of electronic dosimeters in high noise areas; dosimetry placement; airborne radioactivity monitoring; controls for highly activated or contaminated materials (nonfuel) stored within spent fuel and other storage pools; and posting and physical controls for high radiation areas and very high radiation areas

• Radiation worker and radiation protection technician performance with respect to radiation protection work requirements

• Audits, self-assessments, and corrective action documents related to radiological hazard assessment and exposure controls since the last inspection

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

These activities constitute completion of the one

(1) required sample as defined in Inspection Procedure 71124.01-05.

b. Findings

Introduction.

The inspectors reviewed a Green, self-revealing, noncited violation of Technical Specification 5.8.1 for failure to follow radiation work permit requirements.

Two individuals became contaminated while cleaning the gasket seating surface on the endbell of the letdown heat exchanger because they did not use face shields as required by the radiation work permit.

Description.

On November 13, 2009, two individuals alarmed the radiologically controlled area exit contamination monitors after performing flange cleaning activities on letdown heat exchanger CH-7, located in Room 12 of the auxiliary building. The subsequent investigation revealed that the individuals were using scouring pads and a water-glycerin solution to clean the gasket seating surface on the endbell. This work was controlled using Task 4 of Radiation Work Permit 09-3505, CH-7 Endbell Gasket Replacement and Associated Tasks. Task 4 required the use of face shields for abrasive work on exposed system internals. The licensees investigation confirmed that face shields were not used during the cleaning of the gasket seating surface. The air sample taken during the gasket surface cleaning showed airborne activity at seven times the derived air concentration. The licensee immediately restricted the two individuals from entry into the radiologically controlled area, conducted a coaching session with the individuals involved, and placed this issue into the corrective action program as Condition Report 2009-5688. Decontamination methods were successful in removing the contamination from the individuals faces. Subsequent whole body counts indentified low levels of internal contamination with Co-58.

Analysis.

The failure to follow the instructions listed on a radiation work permit was a performance deficiency. The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to follow radiation work permit instructions increased personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because:

(1) it was not associated with ALARA planning or work controls,

(2) there was no overexposure,

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

(4) the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work practices, human error prevention techniques, because the individuals failed to use self-and peer-checking to ensure they were signed onto the appropriate task for the work to be performed and follow the work instructions H.4(a).

Enforcement.

Technical Specification 5.8.1 states that written procedures shall be established, implemented and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, 1978. Section 7.e(1)of Appendix A of Regulatory Guide 1.33 lists radiation protection procedures and access controls to radiation areas, including a radiation work permit system. The licensees Radiation Protection Plan, RPP, Revision 24, Section 2.3.11, states that station personnel are responsible for complying with the requirements in Procedure SO-G-101, Radiation Worker Practices. Procedure SO-G-101, Revision 33, Section 5.5.4, states that personnel signed onto a radiation work permit shall understand the requirements of the radiation work permit. The section also states that personnel shall adhere to the requirements and instructions listed on the radiation work permit. Task 4 of Radiation Work Permit 09-3505 states, in part, that a disposable face shield mask is required for abrasive work on exposed system internals. Contrary to these requirements, on November 13, 2009, two individuals entered Room 12 of the auxiliary building and cleaned the gasket seating surface of the letdown heat exchanger without the use of face shields as required by the radiation work permit, resulting in facial contamination on both individuals. Decontamination methods were successful in removing the contamination from their faces. Subsequent whole body counts identified low levels of Co-58 contamination. Since this violation was of very low safety significance and was documented in the licensees corrective action program as Condition Report 2009-5688, it is being treated as a noncited violation, consistent with Section 2.3.1.a of the NRC Enforcement policy: NCV 05000285/2010004-06, Failure to Follow Radiation Work Permit Requirements.

2RS0 2 Occupational ALARA Planning and Controls

a. Inspection Scope

This area was inspected to assess performance with respect to maintaining occupational individual and collective radiation exposures ALARA. The inspectors used the requirements in 10 CFR Part 20, the technical specifications, and the licensees procedures required by technical specifications as criteria for determining compliance.

During the inspection, the inspectors interviewed licensee personnel and reviewed the following items:

• Site-specific ALARA procedures and collective exposure history, including the current 3-year rolling average, site-specific trends in collective exposures, and source-term measurements

• ALARA work activity evaluations/postjob reviews, exposure estimates, and exposure mitigation requirements

• The methodology for estimating work activity exposures, the intended dose outcome, the accuracy of dose rate and man-hour estimates, and intended versus actual work activity doses and the reasons for any inconsistencies

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

• Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas

• Audits, self-assessments, and corrective action documents related to ALARA planning and controls since the last inspection

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

These activities constitute completion of the one

(1) required sample as defined in Inspection Procedure 71124.02-05.

b. Findings

Introduction.

The inspectors reviewed a Green, self-revealing, noncited violation of Technical Specification 5.8.1 for failure to appropriately control radiation exposures due to improperly planned maintenance activities.

Description.

On June 28, 2010, the inspectors reviewed valve maintenance Work Package 09-AP-20, Valve Maintenance, developed to support the valve maintenance activities, which had an initial dose estimate of 14.084 rem and 8,000 man-hours. The contractor associated with valve maintenance activities planned the various tasks associated with the valve maintenance, valve modifications, and boric acid cleanups.

However, unplanned or unintended dose was received due to inadequate planning and coordination of work activities.

The valve maintenance tasks accumulated an actual total dose of 24.005 rem and 8,975 man-hours. Accounting for dose of 2.138 rem associated with emergent work, the valve work ended with a revised 21.867 rem, which exceeded the initial dose estimate by 7.783 rem, nearly 55 percent. The primary reasons for exceeding the original dose

estimate were improper planning and inadequate oversight of the vendor by the licensee. Review of the condition reports and discussion with licensee personnel indicated that the licensee did not readily control the unintended dose until the dose goal was exceeded. This made the self-revealing aspect of the finding more apparent.

There were numerous examples of improper planning of work activities associated with Work Package 09-AP-20. Walkdowns performed prior to the scheduled valve maintenance activities failed to identify parts and equipment required for the valve work, such as rigging points. As a result, specialty wrenches, scaffold, and other tools were not readily available to the maintenance crew when needed. The licensee identified that 80 percent of the scaffold should have been anticipated on walk-downs. Planning deficiencies also resulted in the failure to anticipate system configuration during execution of valve work, such as identifying the system piping being full or empty which impacted the actual dose received. In one instance, miscommunication resulted in the wrong valve being cut out of the system requiring the job to be reworked, accounting for more unintended dose. Additionally, the licensee did not have adequate interface with the contractor to ensure that procedures were properly reviewed, including ALARA planning.

Analysis.

The failure to properly plan maintenance activities to minimize personnel radiation dose is a performance deficiency. This finding is greater than minor because it affected the Occupational Radiation Safety Cornerstone attribute of program and process in that ALARA planning or radiological controls did not prevent unplanned, unintended dose for a work activity. This caused increased collective radiation dose of for the job activity to exceed the planned dose, approximately 14 rem, by more than 50 percent. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined this finding to be of very low safety significance because the finding involved ALARA planning and controls, and the licensees latest rolling 3-year average does not exceed 135 person-rem. This finding had an associated human performance crosscutting aspect in the work practices component because the licensee did not ensure supervisory and management oversight of work activities, including the contractor, to maintain doses ALARA H.4(c).

Enforcement.

Technical Specification 5.8.1 requires written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, February 1978. Section 9.e(2), Procedures for Performing Maintenance, of Appendix A to Regulatory Guide 1.33 requires, in part, that procedures performed for maintenance be properly preplanned appropriate to the circumstances to minimize radiation exposure. Step 6.1.1 of Procedure RP-AD-300, ALARA Program, Revision 22, states, in part, that the radiation protection department will engage in the work planning process, with the assistance of craft personnel and work supervisors, to minimize the radiation exposure received for the job. Contrary to this requirement, in June 2009, radiation protections involvement with the planning for Work Package 09-A-20 was inadequate, resulting in approximately 7.8 person-rem of unplanned radiation exposure. Since this violation is of very low safety significance and was entered into the corrective action program as Condition Reports 2009-6171,

2009-6264, and 2010-1696, this violation is being treated as a noncited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000285/2010004-07; Failure to Properly Plan a Maintenance Activity.

2RS0 6 Radioactive Gaseous and Liquid Effluent Treatment

a. Inspection Scope

This area was inspected to:

(1) ensure the gaseous and liquid effluent processing systems are maintained so radiological discharges are properly mitigated, monitored, and evaluated with respect to public exposure;

(2) ensure abnormal radioactive gaseous or liquid discharges and conditions, when effluent radiation monitors are out-of-service, are controlled in accordance with the applicable regulatory requirements and licensee procedures;

(3) verify the licensees quality control program ensures the radioactive effluent sampling and analysis requirements are satisfied so discharges of radioactive materials are adequately quantified and evaluated; and

(4) verify the adequacy of public dose projections resulting from radioactive effluent discharges. The inspectors used the requirements in 10 CFR Part 20; 10 CFR Part 50, Appendices A and I; 40 CFR Part 190; the Offsite Dose Calculation Manual, and licensee procedures required by the Technical Specifications as criteria for determining compliance. The inspectors interviewed licensee personnel and reviewed and/or observed the following items:

• Radiological effluent release reports since the previous inspection and reports related to the effluent program issued since the previous inspection, if any

• Effluent program implementing procedures, including sampling, monitor setpoint determinations and dose calculations

• Equipment configuration and flow paths of selected gaseous and liquid discharge system components, filtered ventilation system material condition, and significant changes to their effluent release points, if any, and associated 10 CFR 50.59 reviews

• Selected portions of the routine processing and discharge of radioactive gaseous and liquid effluents (including sample collection and analysis)

• Controls used to ensure representative sampling and appropriate compensatory sampling

• Results of the inter-laboratory comparison program

• Effluent stack flow rates

• Surveillance test results of technical specification-required ventilation effluent discharge systems since the previous inspection

• Significant changes in reported dose values, if any

• A selection of radioactive liquid and gaseous waste discharge permits

• Part 61 analyses and methods used to determine which isotopes are included in the source term

• Offsite dose calculation manual changes, if any

• Meteorological dispersion and deposition factors

• Latest land use census

• Records of abnormal gaseous or liquid tank discharges, if any

• Groundwater monitoring results

• Changes to the licensees written program for indentifying and controlling contaminated spills/leaks to groundwater, if any

• Identified leakage or spill events and entries made into 10 CFR 50.75 (g)records, if any, and associated evaluations of the extent of the contamination and the radiological source term

• Offsite notifications, and reports of events associated with spills, leaks, or groundwater monitoring results, if any

• Audits, self-assessments, reports, and corrective action documents related to radioactive gaseous and liquid effluent treatment since the last inspection

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

These activities constitute completion of the one

(1) required sample, as defined in Inspection Procedure 71124.06-05.

b. Findings

No findings were identified.

2RS0 7 Radiological Environmental Monitoring Program

a. Inspection Scope

This area was inspected to:

(1) ensure that the radiological environmental monitoring program verifies the impact of radioactive effluent releases to the environment and sufficiently validates the integrity of the radioactive gaseous and liquid effluent release program;

(2) verify that the radiological environmental monitoring program implemented is consistent with the licensees technical specifications and/or offsite dose calculation

manual, and to validate that the radioactive effluent release program meets the design objective contained in Appendix I to 10 CFR Part 50; and

(3) ensure that the radiological environmental monitoring program monitors non-effluent exposure pathways, is based on sound principles and assumptions, and validates that doses to members of the public are within the dose limits of 10 CFR Part 20 and 40 CFR Part 190, as applicable. The inspectors reviewed and/or observed the following items:

• Annual environmental monitoring reports and offsite dose calculation manual

• Selected air sampling and thermoluminescence dosimeter monitoring stations

• Collection and preparation of environmental samples

• Operability, calibration, and maintenance of meteorological instruments

• Selected events documented in the annual environmental monitoring report which involved a missed sample, inoperable sampler, lost thermoluminescence dosimeter, or anomalous measurement

• Selected structures, systems, or components that may contain licensed material and has a credible mechanism for licensed material to reach ground water

• Records required by 10 CFR 50.75(g)

• Significant changes made by the licensee to the offsite dose calculation manual as the result of changes to the land census or sampler station modifications since the last inspection

• Calibration and maintenance records for selected air samplers, composite water samplers, and environmental sample radiation measurement instrumentation

• Inter-laboratory comparison program results

• Audits, self-assessments, reports, and corrective action documents related to the radiological environmental monitoring program since the last inspection

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

These activities constitute completion of the one

(1) required sample as defined in Inspection Procedure 71124.07-05.

b. Findings

No findings were identified.

2RS08 Radioactive Solid Waste Processing, and Radioactive Material Handling, Storage, and Transportation (71124.08)

a. Inspection Scope

This area was inspected to verify the effectiveness of the licensees programs for processing, handling, storage, and transportation of radioactive material. The inspectors used the requirements of 10 CFR Parts 20, 61, and 71 and Department of Transportation regulations contained in 49 CFR Parts 171-180 for determining compliance. The inspectors interviewed licensee personnel and reviewed the following items:

• The solid radioactive waste system description, process control program, and the scope of the licensees audit program

• Control of radioactive waste storage areas including container labeling/marking and monitoring containers for deformation or signs of waste decomposition

• Changes to the liquid and solid waste processing system configuration including a review of waste processing equipment that is not operational or abandoned in place

• Radio-chemical sample analysis results for radioactive waste streams and use of scaling factors and calculations to account for difficult-to-measure radionuclides

• Processes for waste classification including use of scaling factors and 10 CFR Part 61 analysis

• Shipment packaging, surveying, labeling, marking, placarding, vehicle checking, driver instructing, and preparation of the disposal manifest

• Audits, self-assessments, reports, and corrective action reports, radioactive solid waste processing, and radioactive material handling, storage, and transportation performed since the last inspection

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

These activities constitute completion of the one

(1) required sample as defined in Inspection Procedure 71124.08-05.

b. Findings

No findings were identified.

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the performance indicator data submitted by the licensee for the third quarter 2010 performance indicators for any obvious inconsistencies prior to its public release in accordance with Inspection Manual Chapter 0608, Performance Indicator Program.

This review was performed as part of the inspectors normal plant status activities and, as such, did not constitute a separate inspection sample.

b. Findings

No findings were identified.

.2 Safety System Functional Failures (MS05)

a. Inspection Scope

The inspectors sampled licensee submittals for the safety system functional failures performance indicator for the period from the third quarter 2009 through the second quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73." The inspectors reviewed the licensees operator narrative logs, operability assessments, maintenance rule records, maintenance work orders, issue reports, event reports, and NRC integrated inspection reports for the period of July 1, 2009, through June 30, 2010, to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one

(1) safety system functional failures sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.3 Mitigating Systems Performance Index - Emergency ac Power System (MS06)

a. Inspection Scope

The inspectors sampled licensee submittals for the mitigating systems performance index - emergency ac power system performance indicator for the period from the third quarter 2009 through the second quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73." The inspectors reviewed the licensees operator narrative logs, operability assessments, maintenance rule records, maintenance work orders, issue reports, event reports, and NRC integrated inspection reports for the period of July 1, 2009, through June 30, 2010, to validate the accuracy of the submittals. The inspectors reviewed the mitigating systems performance index component risk co-efficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one

(1) mitigating systems performance index emergency ac power system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.4 Mitigating Systems Performance Index - High Pressure Injection Systems (MS07)

a. Inspection Scope

The inspectors sampled the licensees submittals for the mitigating systems performance index - high pressure injection systems performance indicator for the period from the third quarter 2009 through the second quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73." The inspectors reviewed the licensees operator narrative logs, operability assessments, maintenance rule records, maintenance work orders, issue reports, event reports, and NRC integrated inspection reports for the period of July 1, 2009, through June 30, 2010, to validate the accuracy of the submittals. The inspectors reviewed the mitigating systems performance index component risk co-efficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensees issue report database to

determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one

(1) mitigating systems performance index high pressure injection system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.5 Occupational Exposure Control Effectiveness (OR01)

a. Inspection Scope

The inspectors reviewed performance indicator data for the fourth quarter 2009 through the first quarter 2010. The objective of the inspection was to determine the accuracy and completeness of the performance indicator data reported during these periods. The inspectors used the definitions and clarifying notes contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, as criteria for determining whether the licensee was in compliance.

The inspectors reviewed corrective action program records associated with high radiation area (greater than 1 rem/hr) and very high radiation area nonconformance.

The inspectors reviewed radiological, controlled area exit transactions greater than 100 mrems. The inspectors also conducted walkdowns of high radiation areas (greater than 1 rem/hr) and very high radiation area entrances to determine the adequacy of the controls of these areas.

These activities constitute completion of one

(1) occupational exposure control effectiveness sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.6 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual

Radiological Effluent Occurrences (PR01)

a. Inspection Scope

The inspectors reviewed performance indicator data for the fourth quarter 2009 through the first quarter 2010. The objective of the inspection was to determine the accuracy and completeness of the performance indicator data reported during these periods. The inspectors used the definitions and clarifying notes contained in Nuclear Energy Institute 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, as criteria for determining whether the licensee was in compliance.

The inspectors reviewed the licensees corrective action program records and selected individual annual or special reports to identify potential occurrences such as unmonitored, uncontrolled, or improperly calculated effluent releases that may have impacted offsite dose.

These activities constitute completion of one

(1) radiological effluent technical specifications/offsite dose calculation manual radiological effluent occurrences sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

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

4OA2 Identification and Resolution of Problems

.1 Routine Review of Identification and Resolution of Problems

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 corrective action program at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. The inspectors reviewed attributes that included the complete and accurate identification of the problem; the timely correction, commensurate with the safety significance; the evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrence reviews; and the classification, prioritization, focus, and timeliness of corrective actions. Minor issues entered into the licensees corrective action program because of the inspectors observations are included in the attached list of documents reviewed.

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

Introduction.

The inspectors reviewed a self-revealing Green noncited violation of Fort Calhoun Station Technical Specification 5.8.1, for the licensees failure to provide an

adequate maintenance procedure for fire protection system flushing. Specifically, while performing OP-PM-FP-1000 on August 19, 2010, water backed up the VA-87 drain line and spilled onto the east switchgear room floor as well as into Room 19 below.

Description.

On August 19, 2010, at 7:57 p.m., the licensee was performing 1 - Turbine Building Drain Valve Flush and Alarm Test of OP-PM-FP-1000, Quarterly Fire Protection Drain Valve Flush and Alarm Test. Water backed up from condensate drain line connected to VA-87, Switchgear Room A Air Handling Unit, out of the drip pan, and onto the floor of the east switchgear room. As more water collected, it ran under motor control center 1B3C where an opening allowed water to drain down to Room 19. In Room 19, water pooled on top and inside of cable trays in sufficient amounts to wet lengths of cable tray which are mounted above safety related equipment as well as on the floor. This condition was recognized at approximately 10:00 p.m.

A search of the condition reporting system resulted in two other instances where water flowed out of the condensate drain line connected to VA-87 and onto the floor of the east switchgear room. Condition Reports 2008-5704 and 2007-5155 detail these instances.

A search of the operator logs confirmed that Procedure OP-PM-FP-1000 was in progress or had been performed the same day as these condition reports. Procedure OP-PM-FP-1000 does not prevent the back up of water into the east switchgear room.

This condition, if left unchanged, has the potential to render safety related equipment in the east switchgear room and Room 19 inoperable.

Analysis.

The inadequate maintenance procedure is a performance deficiency. This finding is more than minor because if left uncorrected the performance deficiency could have the potential to lead to a more significant safety concern. Specifically the use of Procedure OP-PM-FP-1000 allows the potential wetting of safety related equipment in the east switchgear room and Room 19. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609, Appendix A to determine its significance. Using Attachment 4 of that appendix, the inspectors determined that the finding has very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.

Through conversations with the fire protection system engineer and other licensee members and the fact that similar issues have occurred in the past, the inspectors determined that the primary cause of this finding was the failure to adequately assess the significance of previous condition reports which would have required them to perform a more thorough cause evaluation.

Therefore, this finding has a crosscutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that, the resolutions address causes and extent of conditions, as necessary P.1(c).

Enforcement.

Fort Calhoun Station Technical Specification 5.8.1, requires, in part, that the licensee establish, implement, and maintain written procedures recommended in

Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, which requires procedures for performing maintenance. Contrary to the above, Maintenance Procedure OP-PM-FP-1000, Quarterly Fire Protection Drain Valve Flush and Alarm Test, does not prevent water from backing up and spilling out of the VA-87 drain line onto the floor of the east switchgear room and potentially wetting safety related equipment in this room as well as Room 19 due to the opening under motor control center 1B3C. This violation last occurred on August 19, 2010. This condition has existed since at least December 16, 2007. Because the violation was of very low safety significance and was entered into the licensees corrective action program as Condition Report 2010-4423, the violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000285/2010004-08, Inadequate Maintenance Procedure Results in Water in East Switchgear Room and Room 19.

.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 follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program. The inspectors accomplished this through review of the stations daily corrective action documents.

The inspectors performed these daily reviews as part of their daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings were identified.

.3 Selected Issue Follow-up Inspection

a. Inspection Scope

During a review of items entered in the licensees corrective action program, the inspectors followed up on a corrective action item documenting numerous failures of both trains of control room air conditioning, VA-46A and VA-46B, to verify that the corrective actions are commensurate with the significance of the issue. The inspectors also verified that the licensee is identifying operator work around problems at an appropriate threshold, entering them in the corrective action program, and planning or taking appropriate corrective actions.

These activities constitute completion of one

(1) in-depth problem identification and resolution sample as defined in Inspection Procedure 71152-05.

b. Findings

No findings were identified.

.4 (Closed) Unresolved Item 05000285/2009007-02, Failure to Perform Vendor and

Industry Recommended Testing on Safety-Related and Risk Significant 4160 V and 480 V Circuit Breakers

a. Inspection Scope

The inspectors reviewed condition reports, root cause evaluations, vendor documents, and other documents associated with the approximately 20 breakers that had failed over the last 15 years to resolve Unresolved Item 05000285/2009007-02 associated with inadequate maintenance procedures for 4160 V and 480 V safety-related breakers.

b.

Failure to Perform Vendor and Industry Recommended Testing on Safety-Related and Risk Significant 4160 V and 480 V Circuit Breakers

Introduction.

The inspectors identified a Green noncited violation of Technical Specification 5.8.1(a) for inadequate procedures associated with 4160 V and 480 V safety-related breaker maintenance procedures. The inspectors determined that maintenance procedures used to ensure that 4160 V and 480 V safety-related breakers were being maintained and overhauled in a timely manner were inadequate. The licensee did not have an engineering analysis or technical basis to justify the deviation from vendor or Electric Power Research Institute guidance.

Description.

The inspectors identified that the licensee was not performing the maintenance on the breakers as recommended by the vendor or Electric Power Research Institute guidelines. The licensee had completed a review of its breaker maintenance programs in November 2007, and modified it based on Electric Power Research Institute documents TR-106857-V2 and TR-106857-V3, which are preventive maintenance program bases for low and medium voltage switchgear to assure their proper operation. The licensee only implemented portions of the recommended maintenance program. However, no engineering analysis or technical basis existed to justify the limited changes. The licensee had not been performing all vendors or Electric Power Research Institute recommended tests, inspections, and refurbishments on the breakers since they were installed in 1994.

The inspectors reviewed the licensee's circuit breaker maintenance procedures and records. The inspectors determined that the licensee had not refurbished Asea Brown Boveri 4160 V or General Electric 480 V safety-related and risk significant nonsafety-related circuit breakers within the vendor specified 10-year maximum overhaul periodicity or the Electric Power Research Institute guidance of 12 years. In addition, no engineering basis or evaluation was performed to justify the deviation. The inspectors compared the Electric Power Research Institute guidance and vendor-recommended maintenance requirements against the licensee's maintenance procedures. It was determined that the licensee was not performing some of the recommended activities and they had extended the periodicity of some inspections beyond even the Electric Power Research Institute recommended guidelines. The licensees program for medium

and low voltage switchgear and circuit breakers did not include most of the recommended testing and trending. Specifically, no testing of the operation of the 125 Vdc control circuitry was performed at the voltages postulated to exist at the device terminals during design basis events. Industry standards and Electric Power Research Institute guidance recommend reduced control voltage testing as part of breaker maintenance. Vendor overhaul procedures include reduced control voltage testing on the as-found and as-left control circuit.

Additional recommended testing per the preventative maintenance program basis documents TR-106857-V2 and TR-106857-V3 that were not being performed included:

• Thermography inspections of the breakers and switchgear at recommended periodicity and trending of results

• Measurement of the electrical resistance of coils and relays, trended over time to detect progressive failure of winding insulation and give an indication of the condition of these electrical devices

As a result, the inspectors requested the basis for not performing all of the recommended maintenance activities. The licensee had no documented justification for the reduced maintenance. Additionally, the inspectors found that the licensee had failed to update their in-use guidance when operating experience or new vendor information was issued. The licensee was unable to produce documentation demonstrating recommended maintenance had been performed at the appropriate intervals or an analysis which qualified the practice of extending the maintenance and refurbishment intervals. Therefore, the inspectors were concerned about the reliability of the safety-related and safety significant breakers that had not been overhauled within 10 years.

The licensee stated that the 10-year vendor requirement was based on breakers manufactured and lubricated with petroleum-based grease. The license stated that the Asea Brown Boveri circuit breakers were lubricated with synthetic-based grease, Anderol 757, which does not dry out as fast and extends the useful life of the lubrication.

The licensee cited a May 11, 1995, letter from Asea Brown Boveri Combustion Engineering that implied grease hardening was not an issue with Anderol 757 lubricant.

The inspectors did not agree with this assessment because in 1991, Asea Brown Boveri issued a revision to the HK series 4KV breaker technical manual that called for periodic (10-year) cleaning and lubrication of the breaker operating mechanism with Anderol 757 grease. On April 21, 1995, the NRC issued Information Notice 95-22 informing reactor licensees of problems that could result due to grease hardening in ABB HK series 4KV breakers. Additionally, Fort Calhoun Station had identified in the root cause analysis for Condition Report: 2007-2580, FW-2B Circuit Breaker Demand Failure Revision 1 that the May 11, 1995, letter from Asea Brown Boveri Combustion Engineering was incorrect regarding overhaul practices. The issue was entered into the licensees corrective action program under Condition Report 2009-2306.

Below is a summary of circuit breaker failures that occurred at the Fort Calhoun Station from 2001 through July 2009, which were attributed to inadequate maintenance:

Failure Date Failure Number Voltage Equipment Tag Equipment Description Rotation Cycle

10/02/2003 485 480 BT-1B4B Bus Tie Breaker Outage 04/27/2004 496 4160 FW-5A Heater Drain Pump Monthly 02/07/2005 608 4160 FW-5A Heater Drain Pump Monthly 09/07/2006 640 480 AC-3B CCW Pump Monthly 11/18/2006 654 480 BT-1B3A Bus Tie Breaker Outage 01/25/2007 660 4160 AC-10B, HCV-2851 Raw Water Pump Twice Weekly 02/08/2007 661 4160 AC-10C, HCV-2851 Raw Water Pump Twice Weekly 06/17/2007 678 4160 FW-2B Condensate Pump Monthly 05/22/2008 715 480 BT-1B3A Bus Tie Breaker Outage 02/10/2009 746 480 1B4B MCC Main Feed Outage 02/20/2009 747 480 1B4B MCC Main Feed Outage

Analysis.

The inspectors concluded that the licensee failed to evaluate industry and vendor recommended changes and incorporate the changes into their breaker maintenance procedures. Additionally, the inspectors found that the evidence supporting the vendor recommended 10-year maximum refurbishment periodicity was conclusive.

However, the licensee did not have a proper evaluation to support extending the maintenance frequency. Further, the inspectors did not identify any special or additional preventative maintenance requirements that would provide assurance that the breakers would remain operable beyond the vendor-recommended 10-year maximum overhaul periodicity. The inspectors concluded that each of these items were performance deficiencies.

The inspectors determined that this issue affected the procedure quality attribute for maintenance procedures of the Mitigating System Cornerstone of reactor safety.

Specifically, the issue was more than minor because, if left uncorrected, the failure to incorporate the vendor required maintenance and frequency or fully incorporating Electric Power Research Institute maintenance recommendations for extending the service interval into maintenance procedures for safety related breakers, affected the availability, reliability, and capability of mitigating systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability of safety-related breakers refurbished using the deficient procedures cannot be predicted.

The finding was more than minor because it affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Significance Determination Process Phase 1 Screening Worksheet for the Initiating Events, Mitigating Systems, and Barriers Cornerstones the finding was potentially risk significant for multiple systems. Because the probability of multiple system effects is not effectively addressed by a Phase 2 analysis, a Phase 3 analysis was performed. The analyst determined that, while the licensee failed to perform adequate maintenance on the breakers, the actual failure rate of the breakers was no greater than the theoretical

design failure rate. The finding was determined to be of very low safety significance because the deficiency did not result in any loss of function; the finding was not risk significant due to seismic, flooding, or severe weather initiating event; and because other plant-specific analyses that identify core damage scenarios of concern were not impacted. This finding has a crosscutting aspect in the area of problem identification and resolution because the licensee did not effectively incorporate pertinent industry operating experience into the preventive maintenance programs for the 4160 V and 480 V safety-related and risk significant non-safety-related circuit breakers P.2(b).

Enforcement.

Technical Specification Section 5.8.1(a) requires that procedures be developed and implemented in accordance with Regulatory Guide 1.33. Regulatory Guide 1.33, Section 9, which states, in part, Maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Section 9(b) also states, in part, Preventative maintenance schedules should be developed to specify lubrication and equipment inspection schedules. Contrary to these requirements, from 1994, to May 14, 2009, the licensee's preventative and post-maintenance procedures for safety-related and risk significant 4160 V and 480 V circuit breakers, did not include vendor and Electric Power Research Institute recommended testing for performing:

(1) operability tests of breaker control circuits at the minimum expected control voltage levels postulated to exist at the device terminals during design basis event,

(2) measurement of the electrical resistance of coils and relays, trended over time to detect progressive failure of winding insulation and provide an indication of the condition of these electrical devices,

(3) thermography inspections of circuit breakers, and

(4) 10-year overhaul of the circuit breakers as specified in the vendor manuals.

Since this finding was of very low safety significance and has been entered into the licensees corrective action program as Condition Report 2009-2306, this violation is being treated as a noncited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000285/2010004-09, Failure to perform vendor and industry recommended testing on safety-related and risk significant 4160 V and 480 V circuit breakers.

4OA3 Event Follow-up

.1 (Opened and Closed) Licensee Event Report 05000285/2010-004-00:

Acoustic Monitor Failure Due to Inadequate Barriers for Protection of Cable

Flow Element FE-142, accelerometer for pressurizer safety valve RC-142 flow detection, failed its monthly surveillance test on May 26, 2010. One month earlier, on April 28, 2010, FE-142 had previously failed its monthly surveillance test. The failure of flow element FE-142 on April 28, 2010, was identified in Condition Report 2010-2070.

During the investigation for the May 26, 2010, failure, it was discovered that the direct cause for the failure on April 28, 2010, was the same and had not been corrected. As a result of this discovery, FE-142 was reclassified as inoperable from April 28, 2010, to June 2, 2010, and the technical specification action time was exceeded. The licensee

event report was reviewed by the inspectors, with one associated licensee identified violation, which is summarized in Section 4OA7 of this report. This licensee event report is closed.

.2 (Closed) Licensee Event Report 05000285/2010-002-00:

Failed Feeder Cable Due to Inadequate Procedure Causes Station Shutdown

a. Inspection Scope

On April 8, 2010, a ground alarm for 480 Volt Bus 1B3A was indicating intermittently.

The process of isolating loads on the motor control center required securing power to main feedwater isolation valve HCV-1385. Technical Specification 2.0.1 was entered at 4:22 p.m. due to HCV-1385 being inoperable. At 5:40 p.m., insulation on the supply feeder cables to MCC-3A1 was found to be degraded, and the Phase 2 feeder cable was shorted to ground. A plant shutdown was commenced at 5:40 p.m. per the normal shutdown procedure. At 9:23 p.m., the reactor was manually tripped from 22 percent reactor power per the normal shutdown procedure. All systems functioned properly. At 9:23 p.m., the plant entered Mode 3. At 10:33 p.m., HCV-1385 was manually closed and Technical Specification 2.6.1(1) was exited.

b. Findings

Introduction.

A self-revealing Green noncited violation of Fort Calhoun Station, Technical Specification 5.8.1, occurred for an inadequate procedure for verifying the connection between cable lugs and cables. This inadequacy resulted in the loss of Motor Control Center MCC-3A1 and a subsequent plant shutdown.

Description.

On April 8, 2010, after approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of intermittent ground alarms, a ground locked in on 480 volt bus 1B3A at 5:10 a.m. After eliminating most components on bus 1B3A as the source of the ground, the licensee de-energized HCV-1385, the main feedwater motor-operated isolation valve for reactor coolant loop 2B. This action placed the plant into Technical Specification 2.0.1, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to hot shutdown, since HCV-1385 could not close on a valid steam generator isolation signal without power to the valve. After determining HCV-1385 was not the source of the ground, the licensee determined the ground was located on the supply cable to motor control center MCC-3A1 at 4:26 p.m. A plant shutdown was completed at 9:23 p.m.

when the plant entered Mode 3.

Investigation into the ground revealed that the number 2 phase supply cable to MCC-3A1 was shorted to ground and showed signs of substantial degradation. The direct cause of the ground on bus 1B3A was due to a loose lug connection on the number 2 phase supply cable. This resulted in high resistance between the cable and lug, generating more heat at the connection. Over time, the increased heat allowed the cable insulation to break down to the point that a path was created for the cable conductor to ground.

The cause of the loose lugs was attributed to a maintenance procedure for 480 volt motor control centers, last performed on March 29, 2005. Procedure EM-PM-EX-1100, 480 Volt Motor Control Center Maintenance was used to verify that incoming line cable fasteners were snug tight. The procedure directed maintenance workers to tighten to a specific torque value only if loose fasteners were found. The fasteners required torque values are listed in minimum and maximum values; therefore, it was possible to determine if the connections were adequately torque by only checking connections to be snug tight. Over time, this increased heat allowed the cable insulation to break down to the point that a path was created for the cable conductor to ground.

Analysis.

The inspectors determined that the licensees inadequate maintenance procedure was a performance deficiency. This finding was greater than minor because it was similar to non-minor example 4.b in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that a procedural error caused a reactor trip or other transient. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using 4 of that chapter, the inspectors determined that this finding has a very low safety significance (Green) because it was not a design or qualification deficiency, does not represent an actual loss of safety function nor did it screen as potentially risk significant for external events. Since the finding is not indicative of current licensee performance, there is no crosscutting aspect assigned to this finding.

Enforcement.

Fort Calhoun Station Technical Specification 5.8.1, requires, in part, that the licensee establish and implement written procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, which includes procedures for performing maintenance. Contrary to the above, Maintenance Procedure EM-PM-EX-1100, 480 Volt Motor Control Center Maintenance, does not adequately ensure proper torque values between cable lugs and busses. This inadequacy resulted in cable damage to the feeder cable for MCC-3A1 and a subsequent plant shutdown on April 8, 2010, to repair the damaged cable. This performance deficiency has existed since at least March 29, 2005, the last time the maintenance procedure was performed. Because the violation was of very low safety significance and was entered into the licensee's corrective action program as Condition Report 2010-1704, this violation is being treated as a noncited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000285/2010004-10, Inadequate Maintenance Procedure Results in a Plant Shutdown.

4OA5 Other Activities

.1 (Closed) Temporary Instruction 2515/173, Review of the Implementation of the Industry

Groundwater Protection Voluntary Initiative

a. Inspection Scope

An NRC assessment was performed of the licensees groundwater protection program to determine whether the licensee implemented the voluntary Industry Groundwater Protection Initiative, dated August 2007 (Nuclear Energy Institute 07-07, ADAMS

Accession Number ML072610036). The inspectors interviewed personnel, performed walkdowns of selected areas, and reviewed the following items:

• Records of the site characterization of geology and hydrology

• Evaluations of systems, structures, and or components that contain or could contain licensed material and evaluations of work practices that involved licensed material for which there is a credible mechanism for the licensed material to reach the groundwater

• Implementation of an onsite groundwater monitoring program to monitor for potential licensed radioactive leakage into groundwater

• Procedures for the decision making process for potential remediation of leaks and spills, including consideration of the long-term decommissioning impacts

• Records of leaks and spills recorded, if any, in the licensees decommissioning files in accordance with 10 CFR 50.75(g)

• Licensee briefings of local and state officials on the licensees groundwater protection initiative

• Protocols for notification to the local and state officials, and to the NRC regarding detection of leaks and spills

• Protocols and/or procedures for 30-day reports if an onsite groundwater sample exceeds the criteria in the radiological environmental monitoring program

• Groundwater monitoring results as reported in the annual effluent and/or environmental monitoring report

• Licensee and industry assessments of implementation of the groundwater protection initiative

b. Findings

No findings were identified.

.2 (Closed) Temporary Instruction (TI) 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

An NRC inspection was performed to confirm that the licensee has reported their initial inventories of sealed sources pursuant to 10 CFR 20.2207 and to verify that the National

Source Tracking System database correctly reflects 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

• Reviewed adequacy of licensee maintenance, posting, and labeling of nationally tracked sources

b. Findings

No findings were identified.

.3 (Closed) Temporary Instruction (TI) 2515/180, Inspection of Procedures and Processes

for Managing Fatigue

a. Inspection Scope

The inspectors reviewed the licensees procedures and policies to confirm that the Fitness for Duty program adequately implemented fatigue management requirements for individuals subject to 10 CFR Part 26, Subpart I. The inspectors confirmed that the licensee had procedures in place that described:

• The process to be followed after any individual makes a self-declaration that he or she is not fit to safely and competently perform his or her duties for any part of a working tour as a result of fatigue

• The process for implementing the work hour controls

• The process for conducting fatigue assessments

• Disciplinary actions that may be imposed on an individual following a fatigue assessment, and the conditions and considerations for taking those disciplinary actions

The inspectors reviewed the licensees training program to verify implementation and testing of specified knowledge and abilities specified in 10 CFR 26.203(c)(1) and (c)(2).

The inspectors confirmed that the licensees process for developing the annual Fitness for Duty report include provisions for documenting the summary of instances where work hour controls were waived.

The inspectors also confirmed that the licensee had a process in place to retain the following records for at least 3 years or until the completion of all related legal proceedings, whichever is later:

• Work hours for individuals who are subject to the work hour controls

• Shift schedules and shift cycles of individuals who are subject to the work hour controls

• Waivers and the bases for the waivers

• Work hour reviews

• Fatigue assessments

These activities constitute completion of Temporary Instruction 2515/180, Inspection of Procedures and Processes for Managing Fatigue.

b. Findings

No findings were identified.

.4 (Closed) Unresolved Item 05000285/2005011-05, Intake Structure Design

a. Inspection Scope

The team reviewed actions that the licensee completed to resolve concerns associated with the intake structure unresolved item which was reported in NRC Inspection Report 05000285/2005011. The licensees action was to provide original intake structure design documentation for the seismic, tornado, and barge impact analysis.

The licensee could not provide original design documentation, so licensee staff embarked on a reconstitution effort. The licensee provided new calculations for the seismic design, tornado analysis, and barge impact analysis to the inspectors. The inspectors reviewed these analyses to verify that assumptions associated in these documents were correct. Based on the inspectors review and results documented in NRC Inspection Report 05000285/2009006, the inspectors determined that no further inspection is necessary and the unresolved item is closed.

b. Findings

No findings were identified.

4OA6 Meetings

Exit Meeting Summary

On July 7, 2010, the inspectors presented the results of the radiation safety inspection to Mr. T. Nellenbach, Plant Manager, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On July 29, 2010, the inspectors presented the results of the evaluation of changes, tests, and experiments, and permanent plant modification inspection to Mr. J. Reinhart, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On September 2, 2010, the inspectors presented the results of the radiation safety inspection to Mr. T. Nellenbach, Plant Manager, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On October 18, 2010, the inspectors presented the inspection results to Mr. J. Reinhart, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

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 disposition as noncited violations.

.1 Technical Specification 5.8.1 requires written procedures be established, implemented,

and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation),

Revision 2, February 1978. Section 7.e of Appendix A to Regulatory Guide 1.33 requires radiation protection procedures. Procedure RP-307, Use and Control of Temporary Shielding, Revision 18, Step 5.3 states that, No temporary shielding shall be installed, removed or modified unless authorized. Step 7.4.4.c of this procedure states that, Radiation Protection personnel are NOTIFIED prior to removing shielding. Contrary to these requirements, on December 8, 2009, the containment coordinator removed ten lead shielding blankets hanging on a hand rail without notifying radiation protection. The removal of the blankets increased the dose rate on that side of the railing resulting in increased dose rates. The containment coordinator was counseled by the radiation protection supervisor and the ALARA coordinator. The inspectors determined this finding to be of very low safety significance because:

(1) it did not involve ALARA

planning and controls,

(2) there was no overexposure,

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

(4) the ability to assess dose was not compromised.

This issue was entered into the licensee's corrective action program as Condition Report 2009 6454.

.2 Licensee Event Report 05000285/2010-004 identified that accelerometer flow elements

for both pressurizer safety valves were inoperable from April 28 to June 2, 2010. This condition is prohibited by technical specifications after 7 days, therefore meeting the criteria for a condition prohibited by technical specifications on May 5, 2010. The licensee event report was submitted 17 days later, on August 30, 2010. This is a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(i)(B) for failure to submit a required licensee event report within 60 days of a condition prohibited by technical specifications.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

M. Anderson, Supervisor, Radioactive Waste

S. Baughn, Manager, Nuclear Licensing

R. Beck, Supervisor, System Chemistry

J. Bozarth, Radiation Monitoring System Engineer, System Engineering

E. Breault, Health Physicist

D. Brehm, Supervisor, Radiological Equipment and Dosimetry

C. Cameron, Engineer

G. Cavanaugh, Supervisor, Corrective Action

T. Christensen, Senior Operations Engineer, Operations

P. Christensen, Senior Technician, Radiation Protection

R. Clemens, Division Manager, Nuclear Engineering

M. Cole, Analyst, Chemistry

T. Costanzo, Environmental Specialist, Chemistry

H. Faulhaber, Division Manager, Nuclear Engineering

M. Fern, Manager, System Engineering

M. Frans, Manager, Engineering Programs

J. Goodell, Division Manager, Quality and Performance Improvement

D. Guinn, Supervisor, Nuclear Licensing

A. Hansen, Supervisor, Performance Improvement

J. Herman, Manager, Design Engineering

R. Hodgson, Manager, Radiation Protection

K. Hyde, Supervisor, Design Engineering

T. Jamieson, Supervisor, Radiological Operations

D. Lippy, Nuclear Licensing

A. Lollis, Supervisor, ALARA

E. Matzke, Senior Nuclear Licensing Engineer, Regulatory Compliance

S. Miller, Supervisor, Design Engineering

T. Pilmaier, Manager, Performance Improvement

J. Reinhart, Site Vice President

T. Steckelberg, Health Physicist

C. Sterba, Supervisor, Design Engineering

S. Swearingen, Supervisor, Design Engineering

R. Wescott, Manager, Quality

NRC Personnel

W. Burton, Branch Chief, NRO/DNRL

R. Kellar, Senior Enforcement Specialist

M. Maier, Enforcement Specialist

M. Modes, Senior Reactor Inspector, Region I

B. Pham, Branch Chief, NRR/DLR

D. Loveless, Senior Reactor Analyst

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000285/2010-004 LER Acoustic Monitor Failure Due to Inadequate Barriers for Protection of Cable (Section 4OA3)

05000285/2010004-01 NCV Inadequate Documentation of the Adequacy of Design for the Pumps that Transfer Fuel Oil from Storage Tank FO-10 to FO-1 (Section 1R15)

05000285/2010004-02 NCV Failure to Submit a Required Licensee Event Report (Section 1R15)

05000285/2010004-03 NCV Failure to Update the Updated Safety Analysis Report Solid Waste (Section 1R17)

05000285/2010004-04 NCV Failure to Translate Calculation into Calibration Procedure (Section 1R17)

05000285/2001004-05 NCV Failure to Perform a 10 CFR 50.59 Evaluation (Section 1R20)

05000285/2010004-06 NCV Failure to Follow Radiation Work Permit Requirements (Section 2RS01)

05000285/2010004-07 NCV Failure to Properly Plan a Maintenance Activity (Section 2RS02)

05000285/2010004-08 NCV Inadequate Maintenance Procedure Results in Water in East Switchgear Room and Room 19 (Section 4OA2)

05000285/2010004-09 NCV Failure To Perform Vendor And Industry Recommended Testing On Safety-Related And Risk Significant 4160 V And 480 V Circuit Breakers (Section 4OA2)

05000285/2001004-10 NCV Inadequate Maintenance Procedure Results in Plant Shutdown (Section 4OA3)

Closed

05000285/2005011-05 URI Intake Structure Design (Section 4OA5)

05000285/2009007-02 URI Failure to Perform Vendor and Industry Recommended Testing on Safety Related and Risk Significant 4160 and 480 V Circuit Breakers (Section 4OA2)

05000285/2010003-06 URI Failure to Perform a Proper 50.59 Evaluation

05000285/2010-002-00 LER Failed Feeder Cable Due to Inadequate Procedure Causes Station Shutdown (Section 4OA3)

TI 2515/173

Review of the Implementation of the Industry Groundwater Protection Voluntary Initiative (Section 4OA5)

TI 2515/179

Verification of Licensee Responses to NRC Requirement for Inventories of Materials Tracked in the National Source Tracking System Pursuant to Title 10, CFR Part 20.2207 (Section 4OA5)

TI 2515/180

Inspection of Procedures and Processes for Managing Fatigue (Section 4OA5)