IR 05000244/2005003

IR 05000244-05-003; on 04/01/2005 - 06/30/2005; for R. E. Ginna Nuclear Power Plant; Post Maintenance Testing, Refueling and Other Outage Activities, Public Radiation Safety, Other Activities

ML052030013
Person / Time
Site:	Ginna
Issue date:	07/21/2005
From:	James Trapp Division Reactor Projects I
To:	Korsnick M Rochester Gas & Electric Corp
References
IR-05-003
Download: ML052030013 (42)

Text

July 21, 2005

SUBJECT:

R. E. GINNA NUCLEAR POWER PLANT-NRC INTEGRATED INSPECTION REPORT

Dear Mrs. Korsnick:

On June 30, 2005, the US Nuclear Regulatory Commission (NRC) completed an inspection at your R. E. Ginna facility. The enclosed integrated inspection report documents the inspection findings, which were discussed on July 15, 2005, with you and other members of your staff.

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

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

This report documents four NRC-identified findings of very low safety significance (Green).

Three of these findings were also determined to involve violations of NRC requirements.

However, because of their very low safety significance, and because they were entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs) consistent with Section VI.A of the NRC Enforcement Policy. If you contest any NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at Ginna.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

James M. Trapp, Chief Projects Branch 1 Division of Reactor Projects

Mrs. Mary Docket No.:

50-244 License No.: DPR-18 Enclosure:

Inspection Report 050244/2005003 w/ Attachment: Supplemental Information cc w/encl:

M. J. Wallace, President, Constellation Generation J. M. Heffley, Senior Vice President and Chief Nuclear Officer P. Eddy, Electric Division, NYS Department of Public Service C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law C. W. Fleming, Esquire, Senior Counsel, Constellation Energy Group, Inc.

P. R. Smith, New York State Energy Research and Development Authority J. Spath, Program Director, New York State Energy Research and Development Authority T. Wideman, Director, Wayne County Emergency Management Office M. Meisenzahl, Administrator, Monroe County, Office of Emergency Preparedness T. Judson, Central New York Citizens Awareness Network

Mrs. Mary

SUMMARY OF FINDINGS

IR 05000244/2005-003; 04/01/2005 - 06/30/2005; R. E. Ginna Nuclear Power Plant; Post

Maintenance Testing, Refueling and Other Outage Activities, Public Radiation Safety, Other Activities.

The report covered a 3-month period of inspection by resident inspectors and announced inspections by regional specialists. Three Green non-cited violations (NCVs) and one Green finding were identified. The significance of most findings is indicated by their color (Green,

White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

Cornerstone: Initiating Events

Green.

The inspectors identified a non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, on April 5, 2005 when they noted that Ginna did not implement effective corrective action(s) to ensure seismic supports on various instrumentation sensing lines in the pressurizer enclosure were properly installed. Degraded seismic supports had previously been identified by the NRC during plant walkdowns in February 2004 and November 2001.

This finding is more than minor, because it is associated with the Design Control attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety systems. In accordance with Manual Chapter 0609,

Appendix A, Significance Determination of Reactor Findings for At-Power Situations, the inspectors conducted a Significance Determination Process (SDP) phase 1 screening and determined that the finding is of very low safety significance (Green). The SDP process screened to Green since the degraded condition of the seismic mountings for the pressurizer instrumentation sensing lines did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. Corrective actions were completed under work order 20502000. This finding has a cross-cutting aspect in the problem identification and resolution area with an associated causal factor of inadequate problem identification. (Section 1R20)

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green non-cited violation of 10 CFR 50,

Appendix B, Criterion V, Instructions, Procedures, and Drawings when they noted snubber RHU-36 was removed from the A train of the residual heat iv removal (RHR) system when the system was required to be operable without first performing an engineering analysis as required by procedure IP-IIT-5,

Snubber Inspection and Testing Program.

This finding is more than minor because if left uncorrected, the finding would become a more significant safety concern. This finding, which is under the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and affected the objective to ensure availability of systems that respond to initiating events to prevent undesirable consequences, was determined to be of very low safety significance in accordance with Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors conducted a SDP Phase 1 screening using Checklist 3 of Appendix G and determined the finding to be of very low safety significance (Green) since it did not increase the likelihood of a loss of reactor coolant system (RCS) inventory, did not degrade the ability to terminate a leak path or add RCS inventory when needed, and did not degrade the ability to recover decay heat removal systems once lost. (Section 1R19)

Cornerstone: Barrier Integrity

Green.

A finding was identified by the inspectors that contrary to a commitment outlined in a September 19, 2003, letter to the NRC, Ginna did not perform a bare metal inspection of the lower reactor vessel head during the spring 2005 refueling outage. The performance deficiency associated with this finding was a failure of Ginna to develop adequate inspection procedures for the lower reactor vessel head that could identify pressure boundary leakage. As a result, the ability of Ginna personnel to detect leakage from the lower reactor vessel head could be degraded.

The inspector determined that this finding, which is under the RCS Equipment and Barrier Performance attribute of the Barrier Integrity Cornerstone is more than minor because the failure to develop adequate inspection procedures and evaluation guidance could result in a failure to detect a degraded lower reactor vessel head penetration boundary. In accordance with Manual Chapter 0609,

Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the inspectors conducted an SDP Phase 1 screening and determined that the finding is of very low safety significance (Green). Using the Reactor Safety SDP, this finding screened to Green and was of very low safety significance, since the reactor coolant system pressure boundary was not actually degraded. (Section 4OA5).

Cornerstone: Public Radiation Safety

Green.

The inspectors identified a non-cited violation of Technical Specification S.3.1.1 in the Off-site Dose Calculation Manual (ODCM) because the surveillance frequency for the R-22 radioactive liquid effluent monitor was not met. Ginna personnel immediately scheduled the monitor for calibration on the day following identification of the violation.

v This violation is more than minor because it is associated with the cornerstone attribute of maintaining properly calibrated radioactive effluent monitors and affected the Radiation Safety Cornerstone/Public Radiation Safety Cornerstones objective to ensure the adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The violation is of very low safety significance because, while it did impair Ginnas ability to assess dose, Ginna personnel assessed the doses from effluent releases, and the assessed doses did not exceed the dose values in Appendix I to 10 CFR 50. This finding has a cross-cutting aspect in the problem identification and resolution area with a causal factor of effectiveness of corrective actions (Section 2PS1).

Licensee-Identified Violations

None

REPORT DETAILS

Summary of Plant Status

Ginna began the period in Mode 6 defueled. On April 4, 2005, following completion of refueling activities and tensioning of the reactor vessel head, the plant entered Mode 5. On April 10, 2005, the plant entered Mode 2 to perform physics testing. Following the completion of reactor physics testing on April 11, the plant entered Mode 1 and the turbine was synchronized to the grid. Full power was reached on April 13. Later that day, an Electro Hydraulic (EH) system leak on a fitting to the number four turbine control valve necessitated taking the plant to 75%

power to effect repairs. Once repairs were complete, a power ascension occurred and full power was reached on April 15.

On June 6, 2005, power was briefly reduced to 89% at the request of the local energy distributer, Rochester Gas and Electric (RG&E), because of potential concerns RG&E had with the load-carrying capability of the offsite power distribution network. Once RG&E determined that the load carrying capability of the local grid was adequate, Ginna was returned to 100%

power. The power reduction lasted less than four hours.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

==1R01 Adverse Weather Protection (71111.01 - 1 sample, warm weather preparations)

a. Inspection Scope

==

Using Ginna Procedure A-54.4.2, Hot Weather Seasonal Readiness Walkdown, and the Ginna Updated Final Safety Analysis Report (UFSAR) as a reference, the inspectors reviewed Ginnas preparations for hot weather by walking down plant areas. As part of the walkdown, local area temperatures were checked as well as operability of ventilation and air conditioning cooling systems to ensure that the plant was prepared to handle warm weather.

b. Findings

No findings of significance were identified.

==1R04 Equipment Alignment

a. Inspection Scope

Partial System Walkdowns. (71111.04Q - 3 samples)==

The inspectors used plant technical specifications, Ginna operating procedures, plant piping and instrument drawings (P&ID), and the UFSAR as guidance for conducting partial system walkdowns. The inspection reviewed the alignment of system valves and electrical breakers to ensure proper in-service or standby configurations as described in plant procedures and drawings. During the walkdown, the inspectors evaluated material conditions and general housekeeping of the system and adjacent spaces. The inspectors also verified that operations personnel were following plant technical specifications (TS). The following plant system alignments were reviewed:

• On April 6, 2005, the inspectors completed a walkdown of the low temperature overpressure (LTOP) system when it had been placed into service once the plant entered Mode 5. This system was examined because of its high risk-significance when the plant is in Mode 5 status.

• On April 8, 2005, the inspectors completed a walkdown of the B reactor coolant pump lubricating oil collection system. The condition of the B reactor coolant pump (RCP) lubricating oil collection system was examined because of its importance in preventing containment fires, and because the inspectors noted that the system did not appear to be collecting all of the oil that leaked from the RCP during the previous operational period. Ginnas probabilistic risk analysis indicates that plant fires are the most likely initiating events that could lead to a core damage event.

• On May 10, 2005, the inspectors completed a walkdown of the A emergency diesel generator (EDG) system while the B EDG was out-of-service for surveillance and maintenance activities. The condition of the A EDG was examined because of its high risk-significance.

Complete System Walkdown. (71111.04S - 1 sample)

The inspectors conducted a detailed walkdown of the alignment and condition of the containment recirculation fan cooler system. The containment recirculation fan cooler system was chosen because of the important role it would play to provide pressure reduction in containment if a high-energy line break occurred in containment. In addition to verifying proper system alignment as required by plant TS, the plant UFSAR, and Ginna procedures and drawings, the inspector reviewed system maintenance and action reports. None of the action reports or maintenance work orders indicated the performance/ reliability of the system had declined.

b. Findings

No findings of significance were identified.

==1R05 Fire Protection (71111.05 - 10 samples and

==

71111.05A - 1 sample)

1. Quarterly Inspections

a. Inspection Scope

Using the Ginna Fire Protection Program documents as a guide, the inspectors performed walkdowns of the following fire areas to determine if there was adequate control of transient combustibles and ignition sources. The material condition of fire protection systems, equipment and features, and the material condition of fire barriers were also inspected against industry standards. In addition, the passive fire protection features were inspected, including the ventilation system fire dampers, structural steel fire proofing, and electrical penetration seals. The following plant areas were inspected:

• Containment

• A Battery Room

• B Battery Room

• A Emergency Diesel Generator (EDG) Room

• Charging Pump Room

• EDG A Vault

• EDG B Vault

• Standby Auxiliary Feedwater Building

• Turbine Building Basement

• Transformer Yard

l. Findings

No findings of significance were identified.

2. Fire Brigade Drill

a. Inspection Scope

The inspectors observed a test of the Ginna station fire brigade conducted at 7:00 p.m.

on May 30, 2005. The test involved a simulated fire in the Screenhouse Building. The inspectors verified the fire brigade personnel responded quickly to the fire, and used appropriate personal protective equipment. While combating the fire, the inspectors verified the brigade used proper firefighting techniques, and performed satisfactorily as a team. Following the drill, the inspectors verified that the post-drill critique was thorough.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

a. Inspection Scope

The inspectors evaluated Ginnas internal flood protection measures for the B diesel generator room, the control building air handling room, and the charging pump room.

To perform this evaluation, the inspectors reviewed the Ginna UFSAR and Probabilistic Safety Assessment, and toured the applicable areas. The inspectors also observed portions of a maintenance surveillance activity that examined the condition of the room sump pumps.

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection Activities

a. Inspection Scope

The inspectors observed selected samples of nondestructive examination (NDE)activities in process. Also, the inspectors reviewed selected additional samples of completed NDE and repair/replacement activities. The sample selection was based on the inspection procedure objectives and risk priority of those components and systems where degradation would result in a significant increase in risk of core damage. The observations and documentation reviews were performed to verify the activities were performed in accordance with the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code requirements. The inspectors reviewed a sample of inspection reports and Action Reports (ARs) initiated as a result of problems identified during in-service inspection (ISI) examinations. Also, the inspectors evaluated the effectiveness in the resolution of problems identified during selected ISI activities. The inspectors reviewed Ginnas boric acid corrosion control program walkdowns.

The inspectors reviewed the examination reports for thirteen NDE activities. These samples included two samples of welding activities on a pressure boundary and the package for a repair performed in accordance with the ASME Code during the previous operating cycle.

The inspectors observed the evaluation of video records of visual examination (VT)activities to verify the effectiveness of the examiner, process, and equipment to identify degradation of risk-significant systems, structures, and components and to evaluate the activities for compliance with the requirements of ASME Section XI of the Boiler and Pressure Vessel Code.

The inspectors reviewed the examination reports of ultrasonic (UT) testing of two reactor coolant system (RCS) loop A and loop B welds and a 3" pressurizer spray line weld.

The inspectors reviewed the examination reports of liquid penetrant testing (PT) of two pressurizer surge line welds, a high-pressure safety injection line weld and a 4" RHR line weld. The inspectors reviewed the repair package for a rejectable main steam line weld repair which included samples of radiography, ultrasonic, liquid penetrant, and visual inspection NDE methods. All of these reviews showed satisfactory results.

The inspectors reviewed four samples of NDE evaluations which had been initially rejected and subsequently accepted after evaluation.

The inspectors reviewed report "Ginna Station Steam Generator 05RFO Degradation Assessment Report, Report Number SG-SGDA-05-10, Revision 0, 3/21/05." This report documented the steam generator (SG) degradation mechanisms and the planned inspections for the March 2005 refueling outage.

The inspectors reviewed the composition of the pressurizer nozzle material and verified that Temporary Instruction 2515/160, "Pressurizer Penetration Nozzles and Steam Space Piping Connections in US Pressurized Water Reactors," was not applicable to the Ginna Power Plant.

b. Findings

No findings of significance were identified.

==1R11 Licensed Operator Requalification Program (71111.11 - 1 sample, quarterly)

a. Inspection Scope

==

On May 16, 2005, the inspectors observed a licensed operator simulator scenario. The test observed was scenario ECA3132-04, Steam Generator Tube Rupture. The inspectors reviewed the critical tasks associated with the scenario, observed the operators performance, and observed the post-evaluation critique. The inspectors also reviewed and verified compliance with Ginna procedure OTG-2.2, Simulator Examination Instructions.

b. Findings

No findings of significance were identified.

==1R12 Maintenance Effectiveness (71111.12-

a. Inspection Scope

==

The inspectors evaluated Ginnas work practices and follow-up corrective actions for selected system, structure, or component (SSC) issues to assess the effectiveness of Ginnas maintenance activities. The inspectors reviewed the performance history of those SSCs and assessed Ginnas extent-of-condition determinations for those issues with potential common cause or generic implications to evaluate the adequacy of Ginnas corrective actions. The inspectors reviewed Ginnas problem identification and resolution actions for these issues to evaluate whether Ginna had appropriately monitored, evaluated, and dispositioned the issues in accordance with Ginna procedures and the requirements of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance. In addition, the inspectors reviewed selected SSC classification, performance criteria and goals, and Ginnas corrective actions that were taken or planned, to verify whether the actions were reasonable and appropriate. The following issues were reviewed:

• The service water piping to the charging pump room coolers have experienced several leaks over the past six months. Items reviewed included the timeliness of corrective actions.

• Service water has been leaking into the suction piping of the turbine-driven auxiliary feedwater pump, causing elevated sodium levels in the steam generators during quarterly surveillance tests. Actions taken to mitigate this issue were reviewed.

• Following startup after the refueling outage, temperature detector 401A, which is a hot leg reactor coolant system detector and has repeatedly failed. Repairs were monitored during the outage as a new method was tried with fewer connections required in the installation phase. The plant maintenance and operational responses to this failure were reviewed.

Findings No findings of significance were identified.

2.

a. Inspection Scope

The inspector conducted a review of Ginnas periodic evaluation of the implementation of the maintenance rule required by 10 CFR 50.65 (a)(3). The evaluation covered a period from April 19, 2002 to October 16, 2003. The purpose of this review was to ensure that Ginna effectively assessed its (a)(1) goals, (a)(2) performance criteria, system monitoring, and preventive maintenance activities. The inspector verified that the assessment was completed within the required time period and that industry operating experience was properly utilized. Additionally, the inspector verified that Ginna appropriately balanced equipment reliability with unavailability when planning maintenance activities.

The inspector selected a sample of six risk-significant systems to verify that:

(1) the structures, systems, and components were properly characterized,

(2) goals and performance criteria were appropriate,

(3) corrective action plans were adequate, and

(4) performance was being effectively monitored in accordance with EP-2-P-0168, Rev. 10, "Maintenance Rule Monitoring," and other Ginna maintenance rule procedures.

The following systems were selected for this detailed review:

• Containment Isolation System (System CTS02)

• A Motor Driven Auxiliary Feedwater Pump (System AFS01)

• ESF Actuation System Relay Racks, Train A (System SFS01)

• C Instrument Air Compressor (System IAS02)

• Charging Pumps (System CVCS03)

• Main Control Room HVAC Subsystem (System CBV01)

These systems were in a(1) status or were in a(1) status at some time during the assessment period, or had been recently returned to a(2) from a(1). The inspector reviewed corrective action documents for malfunctions and failures of these systems to determine if:

(1) they had been correctly categorized as functional failures;

(2) they were correctly categorized as maintenance-preventable; and

(3) their performance was adequately monitored to determine if classifying a system as (a)(1) or returning the system to a(2) was appropriate. In order to assess the effectiveness of these corrective action plans, the inspector reviewed the 1st quarter 2005 Ginna Maintenance Rule System Status Report.

Findings No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors evaluated the effectiveness of Ginnas maintenance risk assessments required by paragraph a(4) of 10 CFR 50.65. This inspection included discussions with control room operators and scheduling department personnel regarding the use of Ginnas online risk monitoring software. The inspectors reviewed equipment tracking documentation and daily work schedules, and performed plant tours to gain reasonable assurance that actual plant configuration matched the assessed configuration.

Additionally, the inspectors verified that Ginnas risk management actions, for both planned and/or emergent work, were consistent with those described in procedure IP-PSH-2, "Integrated Work Schedule Risk Management." Risk assessments for the following out-of-service systems, structures, and/ or components were reviewed:

• Troubleshooting of a pressurizer steam space manway steam leak, conducted throughout early May 2005, was reviewed on May 9, 2005.

• Troubleshooting and repair of the main feed regulating valve air operator was reviewed on May 10, 2005.

• Troubleshooting and control of higher than expected vibrations on the number nine main turbine bearing throughout the month of May and June 2005.

• Troubleshooting and repair of the containment radiation gaseous detector R-11, which was conducted on June 26 and 27, 2005.

• The June 10, 2005, troubleshooting that was conducted on electrical bus 17 when the B diesel generator output breaker did not close following the loss of offsite power line 751.

b. Findings

No findings of significance were identified.

1R14 Operator Performance During Non-routine Evolutions and Events

b. Inspection Scope

On June 10, 2005, at 1:51pm, offsite power was lost from circuit 751 when a lightning strike damaged an electrical insulator on the supply line causing circuit 751 to de-energize. The loss of circuit 751 in turn caused power to be lost to electrical busses 16 and 17. In response to the event, the B diesel generator (DG) started and re-energized bus 16. However, the DG output breaker to bus 17 did not close. As a result, the B service water pump, which was operating prior to the event, did not restart, leaving only the C service water pump supplying the entire service water system. This caused service water pressure to decrease from over 60 to less that 25 psi.

Complicating the transient was the loss of service water pump A which failed to start when operators attempted to start it from the main control board. To minimize the heat load on the service water system, as required by station Abnormal Operating Procedures operators shut down non-essential components, and in the case of the B DG, installed alternate cooling to the jacket and lube oil heat exchangers from the fire main system. To recover service water pressure/flow, the control room operators re-energized bus 17 from offsite power line 767, which allowed the B service water pump to be restarted.

The inspectors were on site for the event and responded to the control room. While in the control room the inspectors observed the operators implementing the appropriate abnormal operating procedures including AP-SW.2 Loss of Service Water, and AP-Elec.17/18 Loss of Safeguards Bus 17/18. The inspectors verified that operators had investigated the cause of numerous control room panel alarms that resulted from the loss of circuit 751. Further, the inspectors verified that power was restored to electrical bus 17 in accordance with procedures, and that troubleshooting that was performed on the bus prior to the restoration of power was performed safely.

c. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed operability determinations to verify that the operability of systems important to safety was properly established, that the affected components or systems remained capable of performing their intended safety functions, and that no unrecognized increase in plant or public risk occurred. In addition, the inspectors reviewed the following operability evaluations to determine if system operability was properly justified in accordance with IP-CAP-1.1, Technical Evaluation for Current Operability and Past Operability Determination Worksheet:

• Action Report (AR) 2005-1066 and 1067, Loose Mounting Bolts on Service Water Pumps

• AR 2005-1307, Screenhouse Basement (Circulating Water Pump Area) Floor Cracked and Heaved

• AR 2005-1156, TDAFW Pump Shaft Binding

• AR 2005-1335, Report of Discolored Oil in the A RHR Pump

• AR 2005-1855, Turbine Driven Auxiliary Feedwater Pump Leakage at one Stud Findings No findings of significance were identified.

==1R16 Operator Workarounds (71111.16 - 1 Cumulative Sample and 2 Selected Samples

c. Inspection Scope

==

The inspectors conducted a control room walkdown, interviewed operators, and toured the plant with auxiliary operators to verify longstanding degraded conditions had been assessed as required by A-52.16, "Operator Work-Around & Challenge Control." During the plant and control room walkdowns, the inspectors examined maintenance identification tags, local operating instructions, operator aids, temporary alterations and equipment that had been placed out-of-service. The inspectors also reviewed Ginna-identified operator workarounds and operator challenges, and selected corrective action reports. The inspectors also reviewed the Ginna Quarterly Operations Self Assessment of Aggregate Impact Summary, dated May 16, 2005.

As part of the review, the inspectors evaluated the operators' ability to implement normal, off-normal, and emergency operating procedures with the existing equipment deficiencies. The inspectors also determined whether the functional capability of a system or an operator response to an initiating event would be adversely affected. In addition, the inspectors evaluated the cumulative and synergistic effects of the identified operator workarounds to determine whether there was an effect on multiple mitigating systems. This review constituted one sample of cumulative review of operator workarounds.

As part of the review, the following Ginna-identified degraded conditions received additional review:

• A requirement to locally throttle the residual heat removal (RHR) heat exchanger bypass valves AOVs-624/625 to prevent a loss of suction head to the RHR pumps in the event instrument air was lost.

• The installation of remote air monitoring equipment over an inoperable backdraft damper that is located in the clean side of the intermediate building basement.

b Findings No findings of significance were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors observed portions of post-maintenance testing activities in the field to determine whether the tests were performed in accordance with approved procedures.

The inspectors assessed the tests adequacy by comparing the test methodology to the scope of maintenance work performed. In addition, the inspectors evaluated the test acceptance criteria to verify that the tested components satisfied the applicable design and licensing bases and technical specification requirements. The inspectors reviewed the recorded test data to determine whether the acceptance criteria were satisfied. The following post-maintenance testing activities were reviewed:

• PT-12.1, Emergency Diesel Generator A ( May 3, 2005)

• WO 20401295, Inspect/Repair or Replace Sump Pump Discharge Check Valves (May 2, 2005)

• PT-13.11.2, Diesel Generator B Zone Z-21 Smoke Detection Test (May 17, 2005)

• PT-16Q-T, TDAFW Pump Quarterly Test (April 10, 2005)

• WO 20502613, Unable to Drive Income Drive Unit A Past 80.5 inches Which is Roughly the 5-Path Location (May 19, 2005)

• WO 20502474, Charging Pump Room Sump Pump/Not Working (May 31, 2005)

• WO 20401668, RHU-36 Mechanical Snubber Remove/Reinstall(April 6, 2005)

b. Findings

Introduction.

The inspectors identified a Green non-cited violation of 10 CFR 50 Appendix B Criterion V, Instructions Procedures and Drawings, on April 6, 2005, when the inspector identified that, contrary to the requirements of the Ginna inspection and testing program document, maintenance personnel had removed a mechanical snubber from the residual heat removal (RHR) system piping while that system was required to be operable, before an engineering assessment had been performed. Subsequent investigation by the inspector determined that the applicable maintenance procedures governing snubber removal were inadequate.

Description.

Chapter 3.9.3.3.5 of the Ginna Updated Final Safety Analysis Report (UFSAR) states that snubbers are installed on Seismic Category 1 systems to restrain seismic loads. The Ginna residual heat removal (RHR) system is a seismic category 1 system, and as such it has snubbers installed on various portions of the system to support system operability. On April 6, 2005, Ginna was in Mode 5 with both reactor coolant loops not filled. During this time, RHR system operability was required by Technical Specification (TS) 3.4.8, RCS Loops - Mode 5 Loops Not Filled, which states that two RHR loops shall be operable and one RHR loop shall be in operation. While touring the sub-basement of the Auxiliary Building, the inspector noted that maintenance personnel had removed snubber RHU-36 to perform a functional test without first determining if removal of the snubber would adversely impact operability of the train.

RHU-36 is located on the discharge piping of the A RHR pump and was scheduled to be removed from the system for thirteen hours.

The inspector discussed this observation with engineering personnel and was informed that the applicable Ginna procedures governing snubber testing 40.1, Hydraulic Snubber Removal and Installation, and 40.13, Mechanical Snubber Removal and Reinstallation Procedure, allowed removal of snubbers when the plant was in cold shutdown without performing an engineering review. The inspector noted that removal of snubbers without first performing an engineering assessment was contrary to the Ginna snubber testing program document IP-IIT-5, Snubber Inspection and Testing Program, which stated in step 3.12.1 that All safety-related snubbers... shall be operable in accordance with the applicable system operability requirements...

Analysis.

The deficiency associated with this event was a failure of the Ginna maintenance department to have documented instructions, procedures, or drawings appropriate to the circumstances. Because of this deficiency, contrary to the Ginna snubber testing program document IP-IIT-5, a mechanical snubber was removed from the RHR system, while the system was required to be operable, before an engineering analysis that assessed the significance of the removal had been performed. This finding is more than minor because if left uncorrected, the finding would become a more significant safety concern. This finding, which is under the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and affected the objective to ensure availability of systems that respond to initiating events to prevent undesirable consequences, was determined to be of very low safety significance in accordance with Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors conducted a SDP Phase 1 screening using Checklist 3 of Appendix G and determined the finding to be of very low safety significance (Green)since the finding did not increase the likelihood of a loss of reactor coolant system (RCS) inventory, did not degrade the ability to terminate a leak path or add RCS inventory when needed, and did not degrade the ability to recover decay heat removal systems once lost. Further, a subsequent analysis determined removal of the snubber did not affect operability of the RHR system.

Enforcement.

10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings states, in part that, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances.... Contrary to the above, 40.1, Hydraulic Snubber Removal and Installation and 40.13, Mechanical Snubber Removal and Reinstallation Procedure did not provide adequate guidance for removal of hydraulic and mechanical snubbers.

As a result, during the plant refueling outage, snubber RHU-36 was removed from the A train of the RHR system when the system was required to be operable without first performing an engineering assessment as required by procedure IP-IIT-5, Snubber Inspection and Testing Program. Because this condition has been entered into the Ginna station corrective action program in Action Report 2005-2821, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy:

NCV 05000244/2005003-01, Inadequate Procedural Guidance for Removal of Mechanical and Hydraulic Snubbers.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors conducted several tours of the containment area before the plant entered Mode 3 to ensure that to the maximum extent practicable, outage-related material that could adversely affect the performance of the B sump had been removed.

During the tours, in addition to verifying the cleanliness of the containment, the inspectors also examined the general condition of plant equipment including the adequacy of seismic supports and restraints. The inspectors also examined the B sump to ensure the grating was intact, and to verify the sump did not have any debris that could adversely affect the performance of the RHR pumps. Areas in the auxiliary and intermediate building that are normally high radiation areas when the plant is operating at full power, such as the volume control tank and spent fuel pool filter rooms, were toured to ensure potential degraded conditions had been identified by Ginna personnel.

As the plant startup began, the inspectors monitored select activities in the control room to ensure procedures were followed, and work practices were appropriate. Activities that were monitored included synchronizing the turbine to the electrical grid and portions of reactor physics testing.

During power ascension, and shortly after placing the turbine online, Ginna personnel identified that a valve in the electro-hydraulic system was in the closed vice open position, when the number four turbine control valve did not open as expected. To open the valve, the turbine was taken offline. During this event, the inspectors observed operator actions to take the turbine offline.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, on April 5, 2005 when they noted during a walkdown of the containment that Ginna did not implement effective corrective actions to ensure seismic supports for various instrumentation lines in the pressurizer enclosure were properly installed. Degraded seismic supports had been previously identified by the NRC during plant walkdowns in February 2004 and November 2001.

Description.

During a pre-startup walkdown of the containment, the inspectors noted that several sensing lines for the pressurizer and steam generator level instruments were not adequately mounted to the walls of the pressurizer cubicle. Specifically, several instrument sensing lines were not attached to their supports, and several supports were not attached to the adjacent concrete wall. There were a total of nine locations, some with multiple instrumentation sensing lines that were not attached. The mountings and loose or missing brackets were documented in Action Report 2005-1638, Numerous Loose Missing Tubing Clamps In PZR Cubicle, and subsequently corrected by Ginna under work order 20502000, Adjust/Repair Tubing Supports.

The inspectors noted that there had been two previous occasions where degraded and/or inadequate seismic supports were identified by the NRC during plant walkdowns.

Specifically, during plant walkdowns in February 2004 and November 2001, inspectors identified inadequate seismic supports. In both instances, Green Non-Cited violations had been issued.

Analysis.

The performance deficiency for this finding was a failure of Ginna personnel to implement effective corrective action to ensure equipment is adequately supported and /or not degraded. Corrective actions which Ginna had implemented following the discovery of similar non conforming conditions by the NRC in November 2001 and February 2004 have not been effective.

This finding is more than minor, because it is associated with the Design Control attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety systems. In accordance with Manual Chapter 0609, Appendix A, Significance Determination of Reactor Findings for At-Power Situations, the inspectors conducted a Significance Determination Process (SDP) phase 1 screening and determined that the finding is of very low safety significance (Green). The SDP process screened to Green since the degraded condition of the seismic mountings for the pressurizer instrumentation lines did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. Corrective actions were completed under work order 20502000. This finding has a cross-cutting aspect in the problem identification and resolution area with an associated causal factor of inadequate problem identification.

Enforcement.

10 CFR 50, Appendix B, Criterion XVI, Corrective Action states, in part, that measures shall be established to assure that conditions adverse to quality be promptly identified and corrected. Contrary to 10 CFR 50, Appendix B, Criterion XVI, Ginna did not implement effective corrective action(s) following November 2001 and February 2004 inspections to ensure supports for instrument lines in the pressurizer cubicle were adequately supported, until the deficiency was identified by the NRC during a walkdown of containment in preparation for startup from the 2005 outage. However, because the issue was of very low safety significance, and has been entered into Ginnas corrective action program as Action Report 2005-1638, Numerous Loose, Missing Tubing Clamps in PZR Cubicle and corrected under Work Order 20502000, Adjust/Repair Tubing Supports, this issue is being treated as a NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000244/2005003-02, Instrument lines not adequately supported.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors witnessed the performance and/or reviewed test data for the following six surveillance tests that are associated with selected risk-significant systems, structures, and components (SSCs) to verify that TS were followed, and that acceptance criteria were properly specified. The inspectors also verified that proper test conditions were established as specified in the procedures, that no equipment preconditioning activities occurred, and that acceptance criteria had been met.

• PT-12.6B, Diesel Generator Fuel Oil Transfer Pump B Test ( April 14, 2005)

• 1306, Advanced Digital Feedwater Control System Friday Checks (April 15, 2005)

• 1306, Electrical Friday Checks - Inside the Protected Area (April 22, 2005)

• PT-32A, Reactor Trip Breaker Testing - A Train (June 2, 2005)

• PT-17.7A, Control Room Toxic Gas Monitor Operability Test-A Train (June 2, 2005)

• PT-19, Infrared Thermography on Electrical Components (June 20, 2005).

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications

c. Inspection Scope

The inspectors reviewed the following temporary plant modifications to determine whether the temporary change adversely affected system or support system availability, or adversely affected a function important to plant safety. The inspectors reviewed the associated system design bases, including the UFSAR and TS, and assessed the adequacy of the safety determination screening and evaluation. The inspectors also assessed configuration control of the temporary change by reviewing selected drawings and procedures to verify whether appropriate updates had been made. The inspectors compared the actual installation with the temporary modification documents to determine whether the implemented change was consistent with the approved documented modification. The inspectors reviewed the post-installation test results to verify whether the actual impact of the temporary change had been adequately demonstrated by the test. The temporary modifications were reviewed by the inspectors to verify they were installed in conformance with the instructions contained in procedure IP-DES-3, Temporary Modifications.

• 2005-0010, RTD-1 Terminal Block Change in Rack R1 for Failed Thot Resistance Temperature Detector RTD

• 2005-0015, Turbine Bearing, High Vibration Turbine Trip, Auto-Trip Function Blocked

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

On May 16, 2005, the inspectors observed a licensed operator simulator scenario that included a limited test of the Ginna emergency response plan. Scenario ECA3132-04, Steam Generator Tube Rupture, was observed. During the exercise, the crew successfully classified the event in a timely manner, and the drill was counted as a success in the Ginna Drill/Exercise Performance performance indicator.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Public Radiation Safety

2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems (71122.01-10 samples)

a. Inspection Scope

The inspector reviewed radioactive effluent treatment and monitoring equipment, work activities, practices, and procedural implementation during observations and tours of the facilities and inspected procedures, records, and other program documents to evaluate the effectiveness of Ginnas radioactive gaseous and liquid effluent treatment and monitoring systems with respect to public exposure. This inspection activity represents the completion of 10 samples relative to this inspection area (i.e., inspection procedure sections 02.01.a thru d (1), 02.02.a thru k (8), and 02.03.a thru c (1)) in complete fulfillment of the biennial inspection requirements.

Inspection Planning and In-Office Inspection (02.01.a thru d)

The inspector selectively reviewed the Annual Radiological Effluent Release Reports for 2003 and 2004 to verify that the program was implemented as described in the Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual (RETS/ODCM). The inspector selectively examined the reports for significant changes to the ODCM and to radioactive waste system design and operation, for technical justification and documentation, and for dose consequence to the public. The inspector also selectively reviewed the UFSAR description of the radioactive waste systems.

Onsite Inspection (02.02.a thru k)

On May 17, 18, and 19, the inspector walked down the radiation and flow monitors of the gaseous and liquid release systems to observe the current system configuration with respect to the description in the UFSAR, ongoing activities, and equipment material condition. On May 18, the inspector observed the taking of a grab sample from the high conductivity waste tank in the turbine building which was done in preparation for a radioactive liquid batch release. During this inspection, the inspector examined several radioactive liquid waste release permits, records of continuous gaseous releases, and dose calculations, including monthly, quarterly, and annual doses to members of the public.

The inspector selectively reviewed air cleaning system surveillance test results, records of instrument calibrations, performed since the last inspection, for point-of-discharge effluent radiation monitors, and calibration records for flow measurement devices. The inspector also examined calibration records for counting-room radiation measurement instrumentation associated with effluent monitoring and release activities and the associated quality control records. The inspector also reviewed the results from Ginnas most recent quality assurance surveillances and a self-assessment to determine if Ginna met the requirements of the RETS/ODCM.

Identification and Resolution of Problems (02.03.a thru c)

For a description of the inspection activity in this area, see Section 4OA2 of this report.

Related Activities The inspector performed a selective examination of documents (as listed in the List of Documents Reviewed section) to evaluate the effectiveness of the radioactive gaseous and liquid effluent treatment and monitoring systems with respect to public exposure.

The review in this area was against criteria contained in Subpart D of 10 CFR 20 and Appendices A (Criteria 60 and 64), E, and I to 10 CFR 50, the Plant Technical Specifications, the Offsite Dose Calculation Manual, and the UFSAR.

b. Findings

Introduction.

A Green Non-Cited Violation (NCV) was identified for failure to meet the surveillance requirement for the frequency of the radiation calibration for the liquid effluent radioactivity monitor associated with the high conductivity liquid waste tank.

Description.

Plant Technical Specifications and the Offsite Dose Calculation Manual require that the high conductivity liquid waste tank monitor (R-22) be calibrated at least once per eighteen months +/- 25%. On June 5, 2003, Action Report 2003-1224 identified that the surveillance requirement, for the frequency of the radiation calibration for the liquid effluent radioactivity monitor (R-22) associated with the high conductivity liquid waste tank, had not been met. Subsequently, the high conductivity liquid waste monitor (R-22) was calibrated on June 12, 2003. On May 19, 2005, the inspectors identified that the surveillance requirement, for the frequency of the radiation calibration for the liquid effluent radioactivity monitor (R-22), had not been met again.

Analysis.

Ginna personnel acknowledged that this was a performance deficiency in that a technical specification surveillance requirement had inadvertently not been met. This violation is more than minor because it is associated with the cornerstone attribute of maintaining properly calibrated radioactive effluent monitors and affected the Radiation Safety Cornerstone/Public Radiation Safety cornerstones objective to ensure the adequate protection of public health and safety from exposure to radioactive materials released into the public domain. In addition, if left uncorrected, this concern could potentially result in a more significant safety concern (i.e., a non-conservative response by this effluent monitor). This finding did not meet the criteria for traditional enforcement. Using Manual Chapter 0609, Appendix D, the Public Radiation Safety Significance Determination Process (SDP), it was determined that this violation did not involve the radioactive material control or environmental monitoring program branches but did involve the SDPs radioactive effluent release program branch; further, it was determined that this violation was of very low safety significance because, while it did impair Ginnas ability to assess dose, Ginna did assess the doses from effluent releases, and the assessed doses did not exceed the dose values in Appendix I to 10 CFR 50. Therefore, the finding is Green. Additionally, this finding has a cross-cutting aspect in the problem identification and resolution area with a causal factor of effectiveness of corrective actions in that the corrective actions, identified and implemented in response to Action Report 2003-1224 in June of 2003, did not prevent recurrence of missing the same required surveillance for the same effluent monitor in May of 2005.

Enforcement.

Ginna Technical Specification (TS) 5.5, Programs and Manuals, requires that an Offsite Dose Calculation Manual (ODCM) (5.5.1) and a Radioactive Effluent Controls Program (5.5.4) be established, implemented, and maintained. TS 5.5.4.a requires limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests in accordance with the methodology in the ODCM. ODCM surveillance requirement S.3.1.1 requires that each radioactive liquid effluent monitoring instrumentation channel be demonstrated operable by performance of a channel calibration test at the frequency shown in Table 3.1-2.

Table 3.1-2, Radioactive Liquid Effluent Monitoring Surveillance Requirements, states a frequency requirement of R (i.e., at least once per 18 months) for effluent monitor R-22.

TS Surveillance Requirement (SR) 3.0.2 provides for an extension of the time interval to 1.25 times the stated frequency which is allowed for operational flexibility. Contrary to the above, the high conductivity liquid waste monitor, R-22, after being calibrated on June 12, 2003, was not calibrated again within the required interval of 22.5 months.

Because this failure to meet a surveillance requirement is of very low safety significance and has been entered into the corrective action program (CAP) (action report (AR)2005-2425, Radioactivity Monitor Calibration Frequency Not Met, this violation is being treated as a Non-Cited Violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000244/2005003-03, Failure to meet surveillance frequency for calibration of a radioactive effluent monitor.

4OA2 Identification and Resolution of Problems

a. Inspection Scope

Continuous Review As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"

and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the Ginna corrective action program. This review was accomplished by reviewing paper copies of each condition report (CR), attending daily screening meetings, and accessing Ginnas computerized database.

Semi-Annual Review Further, in an effort to identify trends where Ginna personnel have not implemented effective corrective action to prevent recurrence of equipment performance issues, the inspectors reviewed system health reports and conducted a screening review of all ARs initiated since January 2005. Based upon that initial review, and the inspectors knowledge of the plant, several ARs were selected for additional follow-up. Through review of the ARs, and discussions with personnel in the engineering and operations department, the inspector concluded that Ginna personnel were aware of the performance issues discussed in the ARs, and they had implemented corrective action to resolve the performance issues.

b. Findings

No findings of significance were identified.

2.

a. Inspection Scope

During this inspection, the inspector reviewed Ginnas most recent quality assurance surveillances and one self-assessment to determine if identified problems, related to the radioactive effluent treatment and monitoring program, were entered into the corrective action program for resolution. The inspector reviewed selected corrective action reports, related to the radioactive effluent treatment and monitoring program, and discussed these reports with cognizant personnel to determine if the follow-up activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk.

The inspector selected four issues/action reports (ARs) identified in the Corrective Action Program (CAP) for review (i.e., ARs 2004-1201, -1588, and -3249, and 2005-1639). The issues were associated with an evaluation for potential unmonitored releases, tracking the industrys spent fuel pool leak collection issues, the evaluation of isotopic trends since steam generator replacement, and an unknown activity input into a monitor tank.

The documented reports for the issues were reviewed to determine whether the full extent of the issues were identified, appropriate evaluations were performed, and appropriate corrective actions were specified and prioritized.

b. Findings

No findings of significance were identified.

3.

a. Inspection Scope

The inspectors reviewed the corrective action reports listed in Attachment 3, which involved in-service inspection-related problems, to ensure that these issues were properly addressed.

b. Findings

No findings of significance were identified.

4. Identification and Resolution of Problems - (71152-1 sample)

a. Inspection Scope

Since 2002, the boric acid and reactor makeup water flow counters located on the main control board have failed to operate correctly on several occasions. Problems that occurred included failure of the counters to correctly tabulate the amount of boric acid that had been added to the reactor coolant system, to not adding the operator-requested amount of makeup water. The inspector reviewed Ginnas actions to identify, evaluate and correct this problem to ensure it was consistent with the requirements identified in the corrective action program. Additionally, the inspector reviewed Ginnas follow-up corrective actions, which included installation of new totalizers during the 2005 outage, to ensure that long-term corrective action was effective.

b. Findings

No findings of significance were identified.

5. Cross-References to PI&R Findings Documented Elsewhere

Section 1R20 describes a finding for failure to identify numerous deficient mounting brackets and clamping devices for instrument sensing lines in the pressurizer cubicle.

The causal factor of problem identification in the cross-cutting area of problem identification and resolution has been assigned to this finding.

Section 2PS1 describes a finding for failure to meet a surveillance requirement for the frequency of the radiation calibration for a liquid effluent radioactivity monitor. The fact that Ginna had not been properly testing this monitor had been previously identified as an issue by an NRC inspector in June 2003. The causal factor of effectiveness of corrective action in the cross-cutting area of problem identification and resolution has been assigned to this finding.

4OA3 Event Follow-up

1. (Closed) LER 05000244/2005001-00, Failure of ADFCS Power Supplies Results in Plant

Trip On February 16, 2005, the reactor automatically tripped from 100% power. The reactor trip was the result of a turbine trip caused by an anticipated-transient-without-scram (ATWS) mitigation actuation circuitry (AMSAC) signal. The AMSAC signal was generated by low feedwater flow signals, caused by the failure of redundant advanced digital feedwater control system (ADFCS) power supplies. All safety systems operated as designed on the trip with the exception of the main steam atmospheric relief valves (ARV) which were also affected by the failed power supplies. Local manual operation of the ARVs was available. Ginna corrective action included replacing both ADFCS power supplies with station spares and restoring the system to normal operation. The ADFCS cooling muffin fan array was also replaced.

A Ginna investigation concluded the power supplies, which were original equipment installed in 1989, failed because of age-related degradation and overheating. Ginna is developing a program to examine other power supplies that rely on muffin fans for cooling and evaluating the possibility of creating a preventive maintenance program for these elements in the plant. The LER was reviewed by the inspectors and no findings of significance were identified. Ginna documented the failed equipment in Action Report 2005-0624. This LER is closed.

2. (Closed) LER 05000244/2005002-00, Emergency Diesel Generator Start Resulting from

Loss of Off-Site Power Circuit 751 On April 1, 2005, a faulted lightning arrester resulted in the loss of off-site circuit 751.

The site electrical system was in a 50/50 line-up, at the time, causing a loss of safeguards busses 16 and 17. The B emergency diesel generator (EDG) started automatically and re-energized the safeguards busses as designed. The plant was in Mode 6 with refueling in progress. Refueling operations were immediately halted. Core cooling was momentarily interrupted and was promptly restored when the safeguard bus was re-energized. The operating spent fuel pool cooling loop was being powered from the opposite train and was not affected. Off-site power was realigned and all plant loads were returned to off-site power on the redundant source (Circuit 767). The lighting arrester was considered a run-to-failure component by the local electric power supplier Rochester Gas and Electric (RG&E). In the future, RG&E has agreed to conduct periodic diagnostic testing on lightning arresters associated with Circuit 751 and 767.

The LER was reviewed by the inspectors and no findings of significance were identified.

This LER is closed.

4OA5 Other Activities (OA)

1. TI 2515/152 - Reactor Pressure Vessel (RPV) Lower Head Penetration (LHP) Nozzles

(NRC BULLETIN 2003-02)

a. Inspection Scope

The inspectors reviewed Ginnas response to NRC Bulletin 2003-02, which described the RPV lower head penetration (LHP) inspection program. The inspectors reviewed the LHP nozzle examination procedure to determine whether it provided adequate guidance and examination criteria to implement Ginnas examination plan. The inspectors reviewed examination personnel training and qualification records to ensure that personnel were adequately prepared to perform the assigned examination activities.

The inspectors observed selected LHP inspection activities and also reviewed photographs and examination reports to determine whether the inspection procedure was effectively implemented. The inspectors observed the review of several penetration nozzles to evaluate the effectiveness of the visual (VT) examination to verify that the penetration intersection location could be fully accessed to perform a 360-degree examination. The inspection was conducted using a video camera which was manually positioned near each LHP nozzle.

The inspectors reviewed the disposition and actions for observations related to oxidation, debris, and dried boron that impacted the viewing of some penetrations on the lower head. The surface of the lower head, the Inconel weld pads around each penetration and approximately 80% of the annulus areas of the bottom mounted instrumentation (BMI) tubes were covered with a gray carboline coating.

The inspectors reviewed the following action reports (ARs) to ensure that Ginnas actions were appropriate:

• AR-2005-1662, Missed License Renewal commitment on 82/182 SS safe end to lower penetration nozzle welds.

• AR-2005-1228, Unacceptable visual exam on socket welds on end of BMI tubes.

• AR-2005-1287, Evaluate coating on LHP nozzle annulus and its effect on boric acid leakage.

• AR-2005-1286, Level 3 evaluator missed indication during review.

• AR-2005-1285, Not all LHP nozzles inspected to VT-1 as planned.

b. Findings

Introduction.

The inspectors identified that contrary to a commitment outlined in a September 19, 2003, letter to the NRC, Ginna did not perform a bare metal inspection of the lower reactor vessel head during the spring 2005 refueling outage. As a result, the ability of Ginna personnel to detect leakage from the lower reactor vessel head could be degraded

Description.

Between March 22 and April 8, 2005, Ginna personnel performed visual inspections of the reactor vessel bottom mounted instrumentation tube to reactor vessel annulus gap area and the general condition of the reactor vessel lower head at Ginna.

The inspections were performed using procedures VT-116, Revision 8, 3/22/05, "Visual Inspection of Reactor Vessel Head," and VT-101, Revision 11, 2/22/05, "Visual Examination Acceptance Criteria." The visual inspections were performed to satisfy Ginna commitments made relative to NRC Bulletin 2003-02, Leakage From Reactor Pressure Vessel Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity. In a September 19, 2003, response to the Bulletin, Ginna committed to perform a bare metal inspection of all lower reactor vessel head penetrations during refuel outages.

The inspector reviewed video tapes of the inspections conducted during the March-April refuel outage and noted that approximately 80% of the bottom mounted instrumentation tubes had a carboline coating material covering the tube annulus to reactor vessel gap.

This material covered and obscured the tube to reactor vessel annular gap and may prevent an effective inspection for potential early signs of leakage from a degraded J-grove weld and/or instrumentation tube. Despite the presence of the coating obscuring the annular gap, and their commitment to perform a bare metal inspection of the lower reactor vessel head, Ginna non destructive examination (NDE) personnel recorded visual inspection results which indicated there was no masking of the annular gap.

The inspector reviewed procedures VT-116, and VT-101 and noted that contrary to the September 19, 2003, response to Bulletin 2003-02, the procedures did not require the inspections to be performed on bare metal.

Analysis.

The inspector determined that the performance deficiency associated with this finding was a failure of Ginna to develop adequate inspection procedures for the lower reactor vessel head that could identify pressure boundary leakage. This finding, which is under the RCS Equipment and Barrier Performance attribute of the Barrier Integrity cornerstone is more than minor because the failure to develop adequate inspection procedures and evaluation guidance could result in a failure to detect a degraded lower reactor vessel head penetration boundary.

In accordance with Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the inspectors conducted an SDP Phase 1 screening and determined that the finding is of very low safety significance (Green). Using the Reactor Safety SDP, this finding screened to Green and was of very low safety significance, since the reactor coolant system pressure boundary was not actually degraded

Enforcement.

No violation of regulatory requirements occurred. The inspectors determined that the finding did not represent a non compliance since commitments are not regulatory requirements.

2. TI 2515/150 - Reactor Pressure Vessel Head and Head Penetration Nozzles (NRC

Order EA-03-009)

a. Inspection Scope

The inspectors reviewed Ginnas response to NRC Order EA-03-009, which described the reactor pressure vessel head and head penetration nozzle inspections. This outage was the first outage after Ginna had replaced the reactor vessel head in 2003. Per NRC order EA-03-009 inspection of the reactor vessel head and head penetration nozzles was not required. However, Ginna did perform a visual examination of the general condition of the top of the reactor vessel head. NDE examinations were not performed on the penetration nozzles. The inspectors reviewed Ginnas reactor pressure vessel head and head penetration nozzle inspection procedure to determine whether it provided adequate guidance and examination criteria to implement the Ginna examination plan. The inspectors reviewed examination personnel training and qualification records to verify that personnel were properly qualified to perform the examination activities.

The inspectors observed Ginnas inspection activities to verify proper performance of the procedure. No indications of significance were identified during this inspection.

Ginna NDE personnel made a video record of this inspection for future reference.

The inspectors reviewed video records of the inspection and observed the data evaluation by the Level III evaluators. The inspectors selected several penetration nozzles to evaluate the effectiveness of the visual (VT) examination to verify that the penetration intersection location could be fully accessed to reliably perform a 360-degree examination of the intersection region. The inspectors verified by observation and review of video tapes that the RPV head was adequately covered during this inspection.

b. Findings

No findings of significance were identified.

3. TI 2515/163, Operational Readiness of Offsite Power

The inspectors performed Temporary Instruction 2515/163, Operational Readiness of Offsite Power. The inspectors collected and reviewed Ginna procedures and supporting information pertaining to the offsite power system specifically relating to the areas of offsite power operability, the maintenance rule (10 CFR 50.65), and the station blackout rule (10 CFR 50.63). The inspectors reviewed this data against the requirements of 10 CFR 50.63; 10 CFR 50.65; 10 CFR 50 Appendix A General Design Criterion 17, Electric Power Systems; and Plant Technical Specifications. This information was forwarded to the office of Nuclear Reactor Regulation (NRR) for further review.

4OA6 Meetings, Including Exit

Periodically during the inspection period, the resident inspector staff and visiting inspectors would meet with Ginna management to discuss inspection results. For example on May 20, 2005, inspectors Jack McFadden and Andrew Rosebrook met with Ms. M. Korsnick, and other members of her staff who acknowledged their inspection results. On July 15, 2005, the resident inspectors presented the final inspection results to Ms. M. Korsnick and other members of her staff who acknowledged the findings. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

S. Adams

Manager of Operations

B. Flynn

Special Projects Manager Ginna Station

R. Gasper

Radiochemistry Foreman

E. Groh

Assistant Operations Manager (Shift)

T. Harding

Senior Licensing Engineer

K. Holmes

Technician, Radiation Protection

J. Hotchkiss

Mechanical Maintenance Manager

G. Jones

Acting Chemistry Director

P. Lewis

ISI Coordinator

T. Marlow

Plant Manager

J. Pacher

Primary Systems and Reactor Engineering Manager

R. Ploof

Scheduling Manager

R. Whalen

Manager Nuclear Engineering Services

G. Wrobel

Nuclear Safety and Licensing Manager

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Opened and Closed

05000244/2005003-01 NCV Failure to develop adequate procedures concerning the testing and maintenance of mechanical and hydraulic snubbers. (Section 1R19)

05000244/2005003-02 NCV Instrument lines not adequately supported.

(Section 1R20)

05000244/2005003-03 NCV Failure to meet surveillance frequency for calibration of an effluent radiation monitor (Section 2PS1)

05000244/2005003-04 FIN A bare metal inspection of the lower reactor vessel head was not performed during the spring 2005 refuel outage. (Section 4OA5)

Closed

05000244/2005001-00 LER Failure of ADFCS Power Supplies Results in Plant Trip

05000244/2005002-00 LER Emergency Diesel Generator Start Resulting from Loss of Off-Site Power Circuit 751

Discussed

NONE