IR 05000244/2007002

Download: ML071350196

Text

May 15, 2007

Mrs. Mary G. KorsnickVice President, R.E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, New York 14519

SUBJECT: R. E. GINNA NUCLEAR POWER PLANT - NRC INTEGRATED INSPECTIONREPORT 05000244/2007002

Dear Mrs. Korsnick:

On March 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspectionat your R. E. Ginna facility. The enclosed integrated inspection report documents the inspection results, which were discussed on April 26, 2007, with David Holm and other members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's 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 two findings of very low safety significance (Green) that were alsodetermined to be violations of NRC requirements. Additionally, one licensee-identified violation, which was determined to be of very low safety significance is listed in this report. Because the violations were of very low safety significance and were entered into your corrective action program (CAP), the NRC is treating these violations as non-cited violations (NCVs) consistent with Section VI.A.1 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 R.E. Ginna Nuclear Power Plant.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, 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 the M. Korsnick2NRC's document system (ADAMS). ADAMS is accessible from the NRC Website athttp://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/Blake D. Welling, Acting ChiefReactor Projects Branch 1 Division of Reactor ProjectsDocket No. 50-244License No. DPR-18

Enclosure:

Inspection Report 05000244/2007002w/

Attachment:

Supplemental Informationcc 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.

M. Balboni, New York State Deputy Secretary for Public Safety 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

SUMMARY OF FINDINGS

...................................................iv

REPORT DETAILS

..........................................................1

REACTOR SAFETY

.........................................................11R01Adverse Weather Protection .......................................1

1R23 Temporary Plant Modifications ....................................12

1EP6Drill Evaluation

OTHER ACTIVITIES

........................................................144OA1Performance Indicator Verification .................................14 4OA2Identification and Resolution of Problems ............................154OA3Event Followup ................................................16 4OA5Other Activities.................................................19 4OA6Meetings, Including Exit..........................................20 4OA7Licensee-Identified Violations.....................................20ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

................................................A-1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

...........................A-1

LIST OF DOCUMENTS REVIEWED

..........................................A-1

LIST OF ACRONYMS

......................................................A-7

iiiSUMMARY

OF [[]]

FINDINGSIR 05000244/2007-002; 01/01/2007 - 03/31/2007; R. E. Ginna Nuclear Power Plant; EventFollowup; Drill Evaluation.The report covered a 3-month period of inspection by resident inspectors and announcedinspections by regional specialists. Two Green findings, both of which were non-cited violations

(NCV), 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

NRC's program for overseeing the safe

operation of commercial nuclear power reactors is described in

NUR [[]]

EG-1649, "Reactor

Oversight Process," Revision 4, dated December

2006.A.NRC -Identified and Self-Revealing FindingsCornerstone: Initiating EventsGreen. A self-revealing

NCV of

10 CFR Part 50, Appendix B, Criterion

III, "DesignControl," was identified because Ginna failed to control the proper design configuration

of installed plant equipment. Specifically, Ginna failed to update records and

procedures reflecting the design requirement for a vent hole to be drilled in the exhaust

port plug for the main steam isolation valve (MSIV) air actuators. As a result, a

replacement actuator was installed during the October 2006 refueling outage on the "B"

MSIV with a solid vent plug. This caused an inadvertent closure of the

MSIV on

March 16, 2007, and resulted in a reactor trip. Ginna replaced the actuator with a

modified version and placed this issue in the corrective action program.The finding is more than minor because it is associated with the design control attributeof the Initiating Events cornerstone, and it adversely affected the cornerstone objective

of limiting the likelihood of those events that upset plant stability during power

operations. Specifically, the closure of "B"

MS [[]]

IV caused a reactor trip with a safety

injection system actuation. The inspectors determined the finding was of very low safety

significance (Green) using a Phase 1 screening of the finding in accordance with IMC 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for

At-Power Situations." The finding screened to Green because it did not contribute to the

likelihood of a primary or secondary system loss-of-coolant-accident (LOCA) initiator, or

to both the likelihood of a reactor trip and the likelihood that mitigation equipment or

functions would not be available. (Section

4OA 3.5)Cornerstone: Emergency PreparednessGreen. The inspectors identified an

NCV of 10 CFR 50.47(b)(15), radiologicalemergency response training, when they noted that the assigned Emergency Response

Organization (ERO) communicators have not been fully trained on all communicator

responsibilities as outlined in Emergency Plan Implementing Procedure (EPIP) 5-7. For

example, since December 2006, contrary to

EP [[]]

IP 5-7, maintenance personnel who were

filling the role of ERO communicator have not been trained to respond to the control

ivroom when medical and fire events have occurred at the station and properly implementtheir communicator duties. Ginna issued a condition report to address the training

deficiency.The inspectors determined that the failure to ensure that control room communicatorswere fully trained on

ERO communicator responsibilities as described in procedure

EPIP

5-7 was more than minor because it was associated with the ERO readiness aspect of

the Emergency Preparedness cornerstone, and it affected the objective to ensure Ginna

is capable of implementing adequate measures to protect the health and safety of the

public in the event of a radiological emergency. The

EP [[]]

SDP was used to assess the

safety significance of this finding related to the non-risk significant planning standard 10

CFR 50.47(b)(15). Based on

IMC 0609 Appendix B, "Emergency Preparedness SDP"

Sheet 1 for the failure to comply with an NRC requirement and the examples provided in

Section 4.15, this finding was determined to be of very low safety significance (Green).

The finding screened to Green, because the individuals were not trained to the

expectations outlined in

EP [[]]

IP 5-7; however, they had received training on their

communicator duties for declared events. This finding has a cross-cutting aspect in the

area of human performance, because Ginna maintenance personnel who were filling the

role of ERO communicator were not fully trained on the roles and responsibilities of the

position as outlined in

EPIP 5-7. (Section 1

EP6)B. Licensee-Identified ViolationA violation of very low safety significance, which was identified by Ginna, has beenreviewed by the inspectors. Corrective actions taken or planned by Ginna have been

entered into Ginna's corrective action program (CAP). The violation and corrective

action(s) are listed in Section

4OA 7 of this report.

REPORT [[]]

DETAILSSummary of Plant StatusGinna began the period at full Rated Thermal Power (RTP). The reactor tripped onJanuary 27, 2007, because of a failure in the plant turbine control system and returned to full

RTP on February 1, 2007. The reactor tripped a second time on March 16, 2007, when the "B"

MSIV inadvertently went closed, and returned to full

RTP on March 19, 2007. The plant

operated at full

RTP for the remainder of the inspection period.1.

REACTO R

SAFET [[]]

YCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01 - 1 site sample) a.Inspection ScopeOn January 26 and 29, 2007, the inspectors toured areas of the plant that containedequipment and systems that could be adversely affected by cold temperatures. Areas

of focus were the intake structure, auxiliary building, the standby auxiliary feedwater

(SAF) pump room, and "A" and "B" battery rooms. During the tour, the inspectors

verified that temperatures in those rooms did not decrease below the values outlined in

the plant Updated Final Safety Analysis Report (UFSAR). The inspectors used Ginna

Procedure A-54.4.1, "Cold Weather Walkdown Procedure," as a guide during plant

walkdowns. b. FindingsNo findings of significance were identified.1R04Equipment Alignment (71111.04).1Partial Walkdown (3 - samples) a.Inspection ScopeThe inspectors used plant technical specifications (TS), Ginna operating procedures,plant piping and instrument drawings (P&IDs), and the

UFS [[]]

AR 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 of the system and adjacent areas. The inspectors also

verified that operations personnel were following plant TS. The following plant system

alignments were reviewed:*On February 7, 2007, the containment purge system was walked down by theinspectors. This system was examined since improper system alignment could

result in a loss of containment integrity.

2Enclosure*On February 13, 2007, the inspectors performed a walkdown of the "A"emergency diesel generator (EDG) system while the "B" EDG was unavailable

for planned monthly maintenance. This system was examined because it would

have been the primary backup power supply in the event of a loss of offsite

power.*On March 28, 2007, the inspectors performed a walkdown of the boric acidtransfer system for reactor reactivity control. The system was walked down

because of the its significance to reactor control. b.FindingsNo findings of significance were identified..2Complete Walkdown (2 - samples) a.Inspection ScopeThe inspectors performed a detailed walkdown of the alignment and condition of thecontainment mini-purge system. The mini-purge system was chosen because proper

operation, testing and maintenance of the system ensure the containment will remain

functional during certain accident scenarios. In addition to verifying proper system

alignment as specified by plant

TS , the plant

UFSAR, and Ginna procedures and

drawings, the inspector reviewed system maintenance and condition reports (CRs).

None of the CRs or maintenance work orders indicated the performance or reliability of

the system had declined.The inspectors performed a detailed walkdown of the alignment and condition of the 125volt direct current (DC) and Class 1E emergency batteries. The DC power system was

chosen because of the important role it plays in accident mitigation in the event of a loss

of offsite power. In addition to verifying proper system alignment as specified by plant

TS , the plant

UFSAR ,

NRC Generic Letters (

GLs) and Information Notices, and Ginna

procedures and drawings, the inspector reviewed system maintenance and CRs. None

of the CRs or maintenance work orders indicated the performance or reliability of the

system had declined. b. FindingsNo findings of significance were identified.

3Enclosure1R05Fire Protection (71111.05).1Quarterly Inspection (11 - samples) a.Inspection ScopeUsing the Ginna Fire Protection Program documents as a guide, the inspectorsperformed 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 conditions 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:*Intermediate Building Basement, (Fire Zone

IBN -1);*Cable Tunnel, (Fire Area

CT);

• Standby Auxiliary Feedwater Pump Building, (Fire Area SAF);

• Technical Support Center, South Corridor, (Fire Zone IS);

• Technical Support Center, Mechanical Equipment Room, (Fire Zone IM);

• Screenhouse Basement, (Fire Zone SH-1);

• Screenhouse Operating Floor, (Fire Zone SH-2);

• Screenhouse Circulating Water Pump Area, (Fire Zone SH-3);

• Water Treatment Room, (Fire Zone

SB -1

WT);

• Control Room, (Fire Area CC); and

• Relay Room, (Fire Zone RR). bFindingsNo findings of significance were identified..2Fire Brigade Drill - Annual Sample (1 - sample) a.Inspection ScopeThe inspectors observed an announced test of the Ginna station fire brigade onMarch 9, 2007. The test involved a simulated fire in two of the four chlorine injection

pumps that are located in the basement of the screenhouse building. The inspectors

verified the fire brigade personnel, which consisted of five auxiliary operators, responded

within the time lines outlined in the Ginna Fire Protection Program Report, and used

personnel protective equipment specified by Ginna fire fighting procedures. The

inspector verified the brigade used proper firefighting techniques and performed

satisfactorily as a team. Following the drill, the inspectors observed the post-drill

critique.

4Enclosure b.Findings No findings of significance were identified.1R06Flood Protection Measures (71111.06 - 1 internal sample) a.Inspection ScopeTo evaluate Ginna's internal flood protection measures, the inspectors reviewed thepreventive maintenance program for the turbine building (TB) condenser pit sump

pumps, and reviewed how Ginna personnel assessed the information contained in NRC

Information Notices (IN) 2003-08, "Potential Flooding Through Unsealed Concrete Floor

Cracks," IN 2005-11, "Internal Flooding/Spraydown of Safety-Related Equipment Due to

Unsealed Equipment Hatch Floor Plugs and/or Floor Drains," and IN 2005-30, "Safe

Shutdown Potentially Challenged by Unanalyzed Internal Flooding Events and

Inadequate Design." The inspectors also reviewed the plant

UFS [[]]

AR and drawings of

the TB floor drain systems, and walked-down the circulating water pump flood mitigating

trip sensors in the screenhouse,

TB , and portions of the

TB floor drain and sump pump

systems. b.FindingsNo findings of significance were identified.1R07Heat Sink Performance (71111.07 - 2 samples) a.Inspection Scope The inspectors reviewed performance tests, periodic cleaning, eddy current inspections,chemical control methods, tube leak monitoring, tube plugging condition, operating

procedures and maintenance practices for the "A" and "B" closed cooling water (CCW)

heat exchangers (HXs) and the "A" and "B"

EDG jacket water and lube oil

HXs. The

purpose of the review was to verify that controls for the selected components conformed

to Ginna's commitments to GL 89-13, "Service Water System Problems Affecting

Safety-Related Equipment."The inspectors compared the inspection results for the heat exchangers to theestablished acceptance criteria to verify that the results were acceptable and that the

heat exchangers operated in accordance with design. The inspectors walked down the

systems, structures, and components. The inspectors reviewed system health reports

and interviewed applicable system engineers. The inspectors also reviewed a sample of

CR s related to the

CCW [[]]

HX s, and

EDG HXs,to ensure that Ginna was appropriately identifying, characterizing, and resolving

problems related to these systems and components within regulatory requirements and

Ginna's commitments.

5Enclosure b.FindingsNo findings of significance were identified.1R11Licensed Operator Requalification Program (71111.11).1Resident Inspector Quarterly Review (1 - sample) a.Inspection ScopeOn January 10, 2007, the inspectors observed a licensed operator simulator scenario.The test observed was scenario

ES 1213-14, "Ejected

RCCA." The inspectors reviewed

the critical tasks associated with the scenario, observed the operators' performance,

and observed the post-evaluation critique by the evaluation group. The inspectors also

reviewed and verified compliance with Ginna procedure OTG-2.2, "Simulator

Examination Instructions," and assessed the Ginna evaluators' compliance with

guidance contained in that document. b.FindingsNo findings of significance were identified..2Biennial Review (1 - sample) a.Inspection ScopeThe following inspection activities were performed using

NUR [[]]

EG-1021, Revision 9,"Operator Licensing Examination Standards for Power Reactors," Inspection Procedure 71111.11, "Licensed Operator Requalification Program," and NRC Manual

Chapter 0609, Appendix I, "Operator Requalification Human Performance SDP,"

CFR 55.46 Simulator Rule (sampling basis) as acceptance criteria. The inspectors reviewed documentation of operating history since the last requalificationprogram inspection. The inspectors also discussed facility operating events with the

resident staff. Documents reviewed included NRC inspection reports, plant

performance insights, and CRs that involved human performance issues for licensed

operators to ensure that operational events were not indicative of possible trainingdeficiencies.The inspectors reviewed two exam sets for both the comprehensive reactor operatorand senior reactor operator biennial written exams, as well as scenarios and job

performance measures (JPMs) administered during this current exam cycle to ensure

the quality of these exams met or exceeded the criteria established in the Examination

Standards and 10 CFR 55.59. During the onsite weeks of this inspection, the inspectors observed the administration ofoperating examinations to operating crews "A" and "C." The operating examinations

6Enclosureconsisted of two to three simulator scenarios for each crew and one set of five

JPM sadministered to each individual. Conformance with Simulator Requirements Specified in 10

CFR 55.46For the site specific simulator, the inspectors observed simulator performance during theconduct of the examinations, and discrepancy reports to verify compliance with the

requirements of 10 CFR 55.46. The inspectors reviewed simulator maintenance, testing

and control procedures and discussed simulator maintenance, testing, configuration

control and machine operation with members of the simulator maintenance staff. A

sample of simulator tests including transients, core performance, computer real time,

and steady state were also reviewed by the inspectors. Inspectors verified that a

sample of completed simulator CRs from the past two-year period effectively addressed

the described issue. For a listing of the specific simulator tests reviewed see the

attachment for list of documents reviewed.Conformance with Operator License ConditionsConformance with operator license conditions was verified by reviewing the followingrecords:*Remediation training records for three individuals and one crew were reviewedfor the past two-year training cycle;*Proficiency watch-standing and reactivation records. A sample of licensedoperator reactivation records were reviewed as well as a random sample of

watch-standing documentation for time on shift to verify currency and

conformance with the requirements of 10 CFR 55; and*Restoration to active license status of two licensed operators.Licensee's Feedback SystemThe inspectors interviewed instructors, training/operations management personnel, andthree operators for feedback regarding the implementation of the licensed operator

requalification program to ensure the requalification program was meeting their needs

and responsive to their noted deficiencies/recommended changes. The inspectors also

confirmed that selected plant and industry events were incorporated into the

requalification program.On February 9, 2007, the inspectors performed an in-office review of Ginnarequalification exam results. These results included the annual operating tests

administered this year. The inspection assessed whether pass rates were consistent

with the guidance of

NRC [[]]

IMC 0609, Appendix I, "Operator Requalification Human

Performance SDP." The inspectors verified that:*Crew failure rate on the dynamic simulator was less than 20 percent.(Failure rate was 14.3 percent);

7Enclosure*Individual failure rate on the dynamic simulator test was less than or equal to20 percent. (Failure rate was 14.3 percent);*Individual failure rate on the walkthrough test (JPMs) was less than or equal to20 percent. (Failure rate was 0.0 percent);*Individual failure rate on the comprehensive biennial written exam was less thanor equal to 20 percent. (Failure rate was 5.7 percent); and*More than 75 percent of the individuals passed all portions of the exam (85.7percent of the individuals passed all portions of the exam).As per Ginna's program, individuals failing any portion of the requalification examinationare put through remediation training and must pass a retest before being returned to

shift. b.FindingsNo findings of significance were identified.1R12Maintenance Effectiveness (71111.12 - 2 samples) a.Inspection ScopeThe inspectors evaluated Ginna's work practices and follow-up corrective actions forselected system, structure, or component (SSC) issues to assess the effectiveness of

Ginna's maintenance activities. The inspectors reviewed the performance history of

those SSCs and assessed Ginna's extent-of-condition determinations for those issues

with potential common cause or generic implications to evaluate the adequacy of

Ginna's corrective actions. The inspectors reviewed Ginna's 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 Ginna's corrective actions that were

taken or planned, to verify whether the actions were reasonable and appropriate. The

following issues were reviewed:*Failure of control room cooling air compressor AKR01A on November 26, 2006,because of a leaking check valve; and*Main battery "A" vital battery monitor maintenance status. b.FindingsNo findings of significance were identified.

8Enclosure1R13Maintenance Risk Assessments and Emergent Work Control (71111.13 - 8 samples) a. Inspection ScopeThe inspectors evaluated the effectiveness of Ginna's maintenance risk assessmentsrequired 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

Ginna's 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 Ginna's 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 SSCs were reviewed:*Maintenance risk associated with a secondary system transient caused when thecondensate makeup valve failed and subsequent repairs to valve 4315.

(January 5, 2007);*Planned maintenance on the "A," "B" and "C" safety injection pumps to testservice water coolant flow to the pump bearings. Testing was performed utilizing

a different methodology to measure flow. (January 11, 2007);*Planned maintenance on the "A"

EDG ,

RSSP-19, Diesel Generator "A" -Auto-Start Undervoltage Logic Test. Initiation of the test caused the equipment

out-of-service monitor to indicate a higher than expected "orange" risk when

initially attempted on February 8, 2007. The test was completed on

March 20, 2007;*Planned maintenance to replace pressure control valve PCV-135(February 14, 2007);*Unplanned maintenance when the radioactive waste system gas analyzer failedon February 20, 2007. The system was restored on February 23, 2007;*Planned maintenance to remove corrosion susceptible elements from secondarysystems found to contain brass fittings and copper piping. A failed brass fitting

was responsible for a small steam leak on a level detector on the 2B main

stream reheater (MSR) on February 20, 2007. Work continued for several

weeks to correct all the issues identified as extent of condition associated with

the initial identified steam leak (Work began the week of February 26, 2007);*Risk associated with an identified failure to properly correct a fire breach permitin the wall separating the intermediate building and the

TB. (March 5, 2007); and*Risk associated with a new version of procedure

PT 2.2Q, "Residual HeatRemoval System - Quarterly," which initially indicated a higher than expected risk

value for quarterly pump testing. (March 26, 2007). b.FindingsNo findings of significance were identified.

9Enclosure1R15Operability Evaluations (71111.15 - 5 samples) a.Inspection ScopeThe inspectors reviewed operability determinations to verify that the operability ofsystems important to safety were 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":*2007-000876, turbine driven auxiliary feedwater (TDAFW) pump rotating withsteam supply valves closed;*2007-001302,

TDA [[]]

FW pump steam supply check valve 3505B puffing;

• 2007-000729, containment spray piping schedule does not match

P& [[]]

ID;

• 2007-002130, main steam non-return check valves failed closure acceptancecriteria per

PT -2.10.15; and*2007-002021,

TM -402B (Delta T

SP 1 module) lag time found out ofspecification. b. FindingsNo findings of significance were identified.1R17Permanent Plant Modifications (71111.17 - 1 sample) a.Inspection ScopeThe inspectors reviewed plant change request (

PCR) 2006-0031, "Control Room RaisedFloor Upgrade" which was installed in phases during the fourth quarter 2006 and first

quarter 2007. The modification updated and rearranged the workstations located in the

control room, which included installing new flat monitors on the control room panels, and

elevating the desks of the Shift Technical Advisor and Control Room Supervisor. New

carpeting and wall treatments were also installed. The inspectors reviewed PCR

2006-0031, and compared the document to Ginna design basis information including the

Ginna Fire Hazards Analysis Report. As part of the review the inspector observed

portions of the modification installation. b.FindingsNo findings of significance were identified.

10Enclosure1R19Post-Maintenance Testing (71111.19 - 8 samples) a.Inspection ScopeThe inspectors observed portions of post-maintenance testing activities in the field todetermine whether the tests were performed in accordance with approved procedures.

The inspectors assessed the test's 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 reviewedthe recorded test data to determine whether the acceptance criteria were satisfied. The

following post-maintenance testing activities were reviewed:*Standby auxiliary feedwater pump "D" breaker replacement retest, white lighttesting in work package and PT-36Q-D, "Standby Auxiliary Feedwater Pump D -

Quarterly," (January 25, 2007);*PT-12.1,

EDG "A" following Fuel Oil Booster Pump Replacement(February 8, 2007);*

PT -31, Charging Pump Inservice Test, Minimum Flow Test followingreplacement of the "A" charging pump drive belt (February 12, 2007);*PT-17.1, Performance Test of Area Radiation Monitors and High Range AreaRadiation Monitors, following repairs to

RM -10B (March 14, 2007);*

PT-60.13A, Control Room Emergency Air Treatment System (CREATS) Heatingand Cooling System Performance Test - Train "A" completed after repairs to

coolant system due to short cycling (March 15, 2007);*CPI-FT-464, Calibration of Steam Generator a Steam Flow Transmitter FT-464,following damaged caused during isolation of an associated condenser pot

during the forced outage March 16-18, 2007 (March 18, 2007);*PT-2.10.5,

MSIV Shutdown Exercising Requirements, conducted followingreplacement of the

MSIV "B" valve operator when the valve failed closed causing

a plant trip on March 16, 2007 (March 18, 2007); and*PT-22.1, Equipment Hatch Door Seal Leakrate Test, conducted following repairsto the inner door (March 29, 2007). b.FindingsNo findings of significance were identified.1R20Refueling and Other Outage Activities (71111.20 - 2 samples, forced outage) .1Reactor Trip Due to Electro-Hydraulic Control System Failure a.Inspection ScopeOn January 27, 2007, the plant tripped as a result of a loss of load transient caused by afailure in the main turbine electro-hydraulic control (EHC) system. The plant response

to the event was similar to previous loss of load events at Ginna, with the

11Enclosurepower-operated relief valves lifting twice to control pressure in the reactor coolantsystem. During the outage, the inspectors reviewed the control of plant shutdown risk

and Ginna post trip followup work activities. Startup preparations and activities were

observed by the inspectors. These inspection processes constituted one sample of this

inspection procedure for a forced outage. b.FindingsNo findings of significance were identified. .2Reactor Trip Due to Main Steam Isolation Valve "B" Failing Closed a.Inspection ScopeOn March 16, 2007, the plant tripped as a result of a valid reactor trip and safetyinjection signal generated when the "B"

MS [[]]

IV went shut while operating at 100 percent

power. During the outage, the inspectors reviewed the control of plant shutdown risk

and the repairs to the

MS [[]]

IV operator. Following the trip, Ginna performed a walkdown

of containment and identified several valve packing leaks and a minor swagelock steam

leak on a steam flow detector, which were repaired. When maintenance activities were

complete, the inspectors observed several plant startup activities. These inspection

processes constituted one sample of this inspection procedure for a forced outage. b.FindingsNo findings of significance were identified.1R22Surveillance Testing (71111.22 - 6 samples) a.Inspection ScopeThe inspectors witnessed the performance and/or reviewed test data for the followingsix surveillance tests that are associated with selected risk-significant 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,

and that no equipment preconditioning activities occurred, and that acceptance criteria

had been met.*PT-16Q-B, Auxiliary Feedwater Pump B Quarterly, (January 4, 2007);*PT-12.2, EDG B (January 17, 2007);

• PT-32B, Reactor Trip Breaker Testing - Train B (January 19, 2007);

• PT-12.3, Security Emergency Diesel Test (February 7, 2007);

• PT-2.2Q, Residual Heat Removal System - Quarterly ("B" Pump Only) (February 9, 2007); and*PT-13, Fire Pump Operation and System Alignment (March 30, 2007).

2Enclosure b.FindingsNo findings of significance were identified.1R23Temporary Plant Modifications (71111.23 - 1 sample) a.Inspection ScopeThe inspectors reviewed the following temporary plant modification to determinewhether 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 modification was reviewed by the inspectors

to verify it was installed in conformance with the instructions contained in procedure

IP -

DES [[-3, "Temporary Modifications."*2005-0021Temporary leak repair to the floor drain pipe leading from theIntermediate Building basement b.FindingsNo findings of significance were identified.Cornerstone: Emergency Preparedness1EP6Drill Evaluation (71114.06 - 1 sample) a.Inspection ScopeOn January 10, 2007, the inspectors observed scenario]]

ES 1213-14, "Ejected

RCCA," alicensed operator simulator scenario that included a limited test of the Ginna emergency

response plan. 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.The inspectors also reviewed issues related to the training, roles and responsibilities ofEmergency Response Organization communicators.

13Enclosure b.FindingsIntroduction: The inspectors identified a Green

NCV of 10

CFR 50.47(b)(15), whichrequires that radiological emergency response training be provided to those who may be

called on to assist in an emergency. Specifically, the station's control room

communicators have not been fully trained on all

ERO communicator responsibilities.Description:

EPIP 5-7, "Emergency Organization," describes the roles andresponsibilities of personnel assigned to the ERO during a declared emergency event,

i.e., Unusual Event or higher at Ginna. One of the individuals in the ERO is the control

room communicator.

EP [[]]

IP 5-7 states that the control room communicator is required to

report to the control room and make the Shift Manager aware of their presence at the

announcement of an emergency (fire, medical, radiation, etc.). Among other duties, the

communicator is tasked with notifying offsite organizations that an event has occurred at

Ginna and coordinating the arrival and movement of emergency response equipment

that has arrived on site to address the casualty. To perform these tasks, the

communicator must proficiently operate specialized communication equipment in the

control room to ensure the offsite agencies are notified in the timely manner. They must

be knowledgeable of how to coordinate the arrival of offsite fire equipment and

ambulances on site. Since December 2006, the control room communicator position

has been staffed by individuals from the maintenance department. Prior to December

2006, the communicator position was staffed by auxiliary operators.Since maintenance personnel have assumed the role of communicator, the inspectorshave noted that maintenance personnel have not been trained or provided guidance on

certain aspects of the communicator position. For example, the inspectors observed

that they have not routinely reported to the control room when medical emergencies and

fire drills have occurred. During these events, "extra" control room personnel that are

not listed in the control room logs as the site communicator, have filled the role of

communicator rather than the designated individual. During a March 9, 2007, fire drill,

although the communicator reported to the control room to participate in the fire drill, the

individual reported only after being summoned by control room personnel. The

inspectors also observed that although the communicators are responsible for

coordinating the arrival of offsite response equipment, they had not received training on

how to perform this task. The difference between the requirements of

EP [[]]

IP 5-7 and

what training the communicators had received was documented in CR 2007-001975,

"Verify Maintenance Personnel are Trained and Proficient in Responsibilities Delineated

in

EPIP 5-7." The inspectors noted that by not routinely participating in drills, the communicators maylose their proficiency because the equipment that is used in declared

ERO events is

identical to the equipment used in fires and medical emergencies. During a subsequent

fire drill, in which the communicator participated, the individual was not able to locate the

procedures and equipment needed to notify offsite organizations during a simulated

event. This issue was documented in CR 2007-002095, "Weakness in Executing

Communicator Duties."

14EnclosureAnalysis: The inspectors determined that the failure to ensure that control roomcommunicators were fully trained on ERO communicator responsibilities as described in

procedure

EPIP 5-7 was more than minor because it was associated with the

ERO

readiness aspect of the Emergency Preparedness cornerstone, and it affected the

objective to ensure Ginna is capable of implementing adequate measures to protect the

health and safety of the public in the event of a radiological emergency. The

EP [[]]

SDP

was used to assess the safety significance of this finding related to the non-risk

significant planning standard

10 CFR 50. 47(b)(15). Based on

IMC 0609 Appendix B,

"Emergency Preparedness

SDP " Sheet 1 for the failure to comply with an

NRC

requirement and the examples provided in Section 4.15, this finding was determined to

be of very low safety significance (Green). The finding screened to Green, because the

individuals were not fully trained to the requirements outlined in

EP [[]]

IP 5-7; however, they

had received training on their communicator duties for declared events. This finding has

a cross-cutting aspect in the area of human performance because Ginna maintenance

personnel who were fulfilling the role of ERO communicator were not properly trained on

the roles and responsibilities of the position as outlined in

EPIP 5-7. Enforcement: 10

CFR 50.47(b)(15) requires that radiological emergency responsetraining be provided to those who may be called on to assist in an emergency. Section

IV of 10

CFR 50, Appendix E, states, in part, that this training shall include emergency

personnel who are involved in fire control and first aid rescue teams.

EP [[]]

IP 5-7,

"Emergency Organization" states that the ERO communicator is tasked with notifying

offsite organizations that an event has occurred at Ginna and coordinating the arrival

and movement of emergency response equipment that has arrived on site to address

the casualty. Contrary to

10 CFR 50.47(b)(15), since December 2006,

ERO communicators have not received training that is required by Section

IV of 10

CFR 50,

Appendix E, so they can meet their duties as outlined in

EP [[]]

IP 5-7. As a result, since

December 2006, maintenance personnel who were filling the role of ERO communicator

have not routinely responded to the control room when medical and fire events have

occurred at the station, and one occasion during a fire drill, the ERO communicator was

not able to implement their portion of the Ginna emergency response plan. Because

this violation was determined to be of very low safety significance and Ginna entered the

deficiency into their corrective action system in

CR 2007-0001975 it is being treated as a

NCV , consistent with section

VI.A. 1 of the

NRC Enforcement Policy. (NCV05000244/2007002-01, Ginna Communicators Not Adequately Trained To

Implement

EPIP 5-7)4.

OTHER [[]]

ACTIVI [[]]

TIES4OA1Performance Indicator Verification (71151 - 2 samples) a.Inspection ScopeUsing the criteria specified in Nuclear Energy Institute (NEI) 99-02, "RegulatoryAssessment Performance Indicator Guideline," Revision 4, the inspectors verified the

completeness and accuracy of the performance indicator data for unplanned scrams per

7,000 critical hours and scrams with loss of normal heat removal. To verify the accuracy

15Enclosureof the data, the inspector reviewed monthly operating reports,

NRC inspection reportsand Ginna event reports issued during calendar year 2006. b.FindingsNo findings of significance were identified.4

OA2Identification and Resolution of Problems .1Continuous Review of Items Entered into the Corrective Action Program a.Inspection ScopeAs 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

CAP. This review was accomplished by reviewing paper copies of

CRs, attending

daily screening meetings, and accessing Ginna's computerized database. b.FindingsNo findings of significance were identified..2Annual

PI&R Sample -

NUS Modules (71152 - 1 sample) a.Inspection ScopeFollowing startup from the 2006 refuel outage, Ginna personnel identified that two signalconditioning modules in the reactor protection overpower differential temperature trip

circuitry were not operating properly. The modules were manufactured by NUS

corporation, and were installed during the refuel outage as part of the extended power

uprate project. An examination of the modules by Ginna and NUS personnel

determined that the failures were caused by abraided wires which had shorted to

ground. The wires were apparently damaged during manufacture of the modules.

These issues were documented in CR-2006-06132, and Ginna's corrective actions

consisted of taping the defective wires and examining the remaining NUS modules that

were installed during the refuel outage. On December 14, 2006, NUS Corporation

informed the

NRC of the manufacturing defect per 10

CFR 21. The inspectors reviewed

the corrective actions outlined in

CR 2006-06132 and several other

CRs relating to the

reliability of the

NUS modules against the requirements of the

CAP to ensure that the

full extent of the issues were identified and that the proposed corrective actions are

appropriate. The inspectors examined two of the defective modules and interviewed

relevant station personnel.

16Enclosure b.Findings and ObservationsNo findings of significance were identified. The corrective action in progress to examinesimilar NUS modules installed in the plant and in stock is appropriate. The corrective

action is being given prioritization and oversight commensurate to its safety and risk

significance.4OA3Event Followup (71153).1Inadvertent Deluge System Initiation a.Inspection Scope (1 - sample)On January 11, 2007, Ginna operations personnel conducting a performance test on thecontrol building air handling room deluge system, SO6, caused fire suppression panels

S22 and S23 to be wetted by backspray from the discharge of the SO6 system. Wetted

panel S22, which controls the deluge systems for 12A off-site power supply transformer

initiated a deluge of the 12A transformer. The inspectors followed Ginna's response to

the event, the corrective actions taken and the compensatory actions put in place while

the wetted panels were removed from service. There was no impact on the plant or the

transformers by the wetting of the S22 and S23 deluge systems. The fire suppression

panels were restored to service on January 13, 2007. b.FindingsNo findings of significance were identified..2Reactor Trip Due to Turbine Control System Malfunction a.Inspection Scope (1 - sample)On January 27, 2007, the Ginna reactor tripped from 100 percent power as a result of avalid over temperature differential temperature signal that was created when all four

turbine control valves closed due to a faulty card in the turbine EHC system. Plant

systems and components responded as designed with both power operated relief

valves, and one steam generator safety valve lifting to relieve excess primary and

secondary plant pressure. Following the reactor plant trip, a minor chemistry transient in

the secondary plant occurred when circulating water entered the feed and condensate

system through a failed check valve in the steam generator blowdown system. This

event was similar to an event that had occurred in July 2005. Following the trip, the

plant was stabilized in Mode 3, "Hot Shutdown," with steam generator level and

pressure being controlled by the steam-driven auxiliary feedwater pump and

atmospheric relief valve systems.The resident inspectors reviewed Ginna's post-trip assessment, and plant restartactivities. An independent post-trip review was performed by the inspectors.

17Enclosure b.FindingsNo findings of significance were identified..3Steam Leak on Main Steam Reheater (MSR) a.Inspection Scope (1 - sample)On February 20, 2007, a brass fitting on the MSR 2B failed, resulting in a steam leakheard by operators in the turbine building. The leak was isolated by operators by

closing two valves locally in the vicinity of the failed component. A subsequent

evaluation of the failed fitting indicated that the part had failed due to excessive

corrosion of the brass fitting in the high pH environment of the secondary systems. The

acceptability of the brass fittings had been questioned previously by plant personnel in

January of 2006, but were not replaced when engineering personnel determined that

pressure rating of the fitting was acceptable. Although the pressure rating was

acceptable, the ability of the fitting to remain operable under the corrosive affects of the

higher pH environment of the secondary plant systems was not considered. The

secondary pH was raised to reduce iron transport to the new steam generators after

their installation in 1996. The failed fitting was likely installed during a major refit of all

the MSR components in 1984. Ginna has subsequently identified and replaced several

brass fittings since the event as part of their extent of condition response to this event. b.FindingsNo findings of significance were identified..4(Closed)

LER 05000244/2006007-01 Main Steam Safety Valve (

MSSV ) SetpointExceedanceThis

LER is an update of an event evaluated in the 4th quarter 2006

NRC inspectionreport 05000244/2006005. That report generated an unresolved item, which was

opened pending Ginna's determination of the root cause(s) for the

MS [[]]

SV failures, and

identification of a possible performance deficiency which may have caused two of eight

main steam relief valves to lift at a pressure above the acceptance band during testing

in October 2006. The

URI is closed in section 4

OA5 of this report. The apparent cause

for the failure of the valves to lift prior to exceeding the limiting pressure was determined

to be increased friction in the spindle guide area of the valve. Ginna is evaluating

whether to refurbish the valves during the next refueling outage. The enforcement

aspects of this issue are discussed in Section

4OA 7 of this report. This

LER is closed..5Reactor Trip Due to Inadvertent

MS [[]]

IV Closure

18Enclosure a.Inspection Scope (1 - sample)On March 16, 2007, the station experienced a reactor trip from 100 percent power as aresult of the "B"

MS [[]]

IV failing shut and causing a valid safety injection signal on low

steam pressure in the "A" steam flow header, with an associated trip and subsequent

MSIV isolation of the "A"

MSIV on high steam flow in the same header. Plant systems

and components responded as designed. No primary power-operated relief valves lifted

during the event. At least one steam generator safety valve on the "B" steam generator

lifted to relieve excess pressure in the isolated "B" steam generator. Following the trip,

the plant was stabilized in Mode 3, "Hot Shutdown," with steam generator level and

pressure being controlled by the motor auxiliary feedwater pumps and atmospheric relief

valves.The resident inspectors reviewed Ginna's post-trip assessment, and plant restartactivities. An independent post-trip review was performed by the inspectors. b.FindingsIntroduction: A self-revealing

NCV of 10

CFR Part 50, Appendix B, Criterion III, "DesignControl," was identified because Ginna failed to control the proper design configuration

of installed plant equipment. Specifically, during the October 2006 refuel outage, an

improperly designed air actuator was installed on the "B"

MSIV. The

MSIV actuator

subsequently failed and caused a reactor trip. Description: Each

MS [[]]

IV at Ginna contains an air actuator that supplies the motive force to open and close the valves. The actuator has two main parts: a spring that holds the

valve closed, and a piston assembly that is used to open the valve. Valve opening is

accomplished when sufficient instrument air has been supplied to the top of the piston to

overcome the force of the closing spring. During the early 1970s, several reactor plant trips occurred at Ginna when air leakagepast the actuator piston seals equalized pressure across the piston allowing the closure

spring to overcome the piston opening force. To correct this condition, a vented plug

was installed in the exhaust side of the actuator in place of a previously installed solid

plug. However, this modification was not identified in plant design drawings, station

procedures or warehouse purchase orders. This oversight did not have immediate

adverse consequences, because Ginna rebuilt the

MS [[]]

IV actuators during planned

maintenance activities, and reused the vented plugs. However, during the October

2006 refuel outage, Ginna changed the previous maintenance practice of rebuilding the

installed operator, and instead installed a replacement item in the "B"

MS [[]]

IV which had

not been modified in accordance with the vendor's directions. As a result, similar to

what occurred in the early 1970's, once sufficient air leakage past the piston seals had

occurred, the "B"

MSIV closed. Ginna replaced the

MSIV vent plug and verified

"A"

MS [[]]

IV was not subject to the same design error.

19EnclosureAnalysis: The performance deficiency is a failure of Ginna to properly control theconfiguration of installed plant equipment, including translation of design information into

plant procedures and work control documents so they reflect the desired plant

configuration. The finding is more than minor because the deficiency is associated with

the design control attribute of the Initiating Events cornerstone, and adversely affected

the cornerstone objective of limiting the likelihood of those events that upset plant

stability during power operations. Specifically, the inadvertent closure of "B"

MS [[]]

IV

resulted in a reactor trip with safety injection system actuation. The finding was

determined to be of very low safety significance (Green) in accordance with IMC 0609,

Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., "Determining the Significance of Reactor Inspection Findings for At-Power

Situations," based on a Phase 1 analysis because the finding did not contribute to the

likelihood of a primary or secondary system

LO [[]]

CA initiator, or to both the likelihood of a

reactor trip and the likelihood that mitigation equipment or functions would not be

available. Enforcement:

10 CFR Part 50, Appendix B, Criterion

III, "Design Control," requires, inpart that measures shall be established to assure that applicable regulatory

requirements and the design basis are correctly translated into specifications, drawings,

procedures, and instructions. Contrary to the above, the "B"

MS [[]]

IV actuator installed

during the outage of October 2006 was not correctly designed because proper design

information was not translated into onsite procedures and drawings associated with this

modification. The

MS [[]]

IV operator subsequently failed shut causing a reactor trip with

safety injection on March 16, 2007. Because this deficiency was determined to be of

very low safety significance and Ginna entered the deficiency into their corrective action

system in

CR 2007-002172, it is being treated as a

NCV , consistent with section

VI.A. 1of the

NRC enforcement policy. (NCV 05000244/2007002-02, Failure of "B"

MSIVD ue to Inadequate Design Control).6(Closed)

LER 05000244/2007001 Loss of Electrical Generation Results in Plant TripThe circumstances involving this incident were previously reviewed in Section 1R20 and4OA3.2 of this report. This

LER is closed.4

OA 5Other Activities(Closed)

URI 05000244/2006005-02, Determination of Performance DeficiencyAssociated with

TS 3.7.3, Main Steam Safety Valves a.Inspection ScopeThis item was opened pending determination of a performance deficiency, which causedtwo main steam relief valves to lift outside of their acceptance band during the October

2006 refuel outage. As outlined in section 4OA3.4 of this report, Ginna determined that

the failure of the valves to lift within the acceptable band was caused by increased

friction in the spindle guide area likely caused by small amounts of grit build up on the

spindle and a manufacturer-identified tendency for bearing material to close up on the

spindle over time. Ginna is evaluating whether to refurbish the valves during the next

refueling outage. This is a licensee-identified finding, and the enforcement aspects of

this violation are discussed in Section

4OA 7 of this report. This

URI is closed.

20Enclosure b.FindingsNo findings of significance were identified.4OA6Meetings, Including ExitExit Meeting SummaryOn April 26, 2007, the resident inspectors presented the inspection results toMr. David Holm and other members of his staff, who acknowledged the findings. Theinspectors asked Ginna's supervision whether any of the material examined during the

inspection should be considered proprietary. No proprietary information was identified.4OA7Licensee-Identified ViolationsThe following violation of very low safety significance (Green) was identified by Ginnaand is a violation of

NRC requirements, which meets the criteria of Section

VI of the

NRC Enforcement Policy,

NUREG -1600, for being dispositioned as a

NCV.TS 3.7.1 requires that all

MSSVs shall be operable. Contrary to this requirement, on

October 7, 2006, with the plant in Mode 1, Ginna performed in-place testing of all

MS [[]]

SVs and determined that two of the valves lifted outside the acceptance band. Since

the two unsatisfactory as-found lift pressures may have arisen over a period of time

(found during sequential testing), Ginna determined that at least one

MS [[]]

SV was

inoperable for longer than the allowed outage time during the previous operating cycle.

This condition was documented by Ginna in CR 2006-004751 and subsequently

corrected. This finding is of very low safety significance because it did not increase the

probability or consequences of a core damage event.ATTACHMENT:

SUPPLE [[]]

MENTAL [[]]

INFORM [[]]

ATION A-1AttachmentSUPPLEMENTAL

INFORM [[]]

ATIONK EY

POINTS [[]]

OF [[]]

CONTAC [[]]

TLicensee personnelM. KorsnickVice President, GinnaD. HolmPlant Manager

J. YoeOperations Manager

D. BlankenshipManager, Radiation Protection

E. GrohAssistant Operations Manager (Shift)

S. KennedyEmergency Preparedness Manager

J. PacherManager, Nuclear Engineering Services

B. RandallNuclear Safety and Licensing Manager

W. ThomsonChemistry Supervisor

R. Whalen Manager, Ginna Maintenance

LIST [[]]

OF [[]]

ITEMS [[]]

OPENED ,

CLOSED ,

AND [[]]

DISCUS SEDOpened and Closed05000244/2007002-01NCVGinna Communicators Not Adequately Trained ToImplement

EPIP 5-705000244/2007002-02

NCVF ailure of "B"

MSIV due to Inadequate DesignControlClosed05000244/2006005-02

URIDetermination of Performance DeficiencyAssociated with TS 3.7.3, Main Steam Safety

Valves05000244/2006007-01LERMain Steam Safety Valve Setpoint Exceedance

05000244/2007001-00LERLoss of Electrical Generation Results in Plant TripLIST

OF [[]]

DOCUME NTS

REVIEW [[]]

EDSection 1R01: Adverse Weather ProtectionER-SC.3, Low Screenhouse Water LevelSection 1R04: Equipment AlignmentProceduresO-11, Control of Mini-Purge Exhaust Valves While Depressurizing ContainmentPT-2.5.1, Air Operated Valves, Quarterly Surveillance Containment

PT-17.2, Area Radiation Monitors R-11 to R-22 and Iodine Monitors R-10 and R-10B

S-23.2, Containment Purge Procedure

A-2AttachmentS-23.2.3, Containment Mini-Purge System OperationProgramsGinna

IST Program DocumentGinna

ISI Program DocumentDrawings33013-1865Containment

HVAC Systems Purge Supply33013-1866Containment

HVAC Systems Purge Supply

33013-1870Auxiliary/Intermediate Building

HV [[]]

AC Systems

202-0102125 VDC Power Distribution System

33013-1239(Sheet 1 of 2) Diesel Generator - A

33013-1266Chemical and Volume Control System Boric Acid (CVCS)DocumentsNRC Generic Letter (GL) 91-06Resolution of Generic Issue A-30, April 29, 1991Ginna Response to

GL 91-06Resolution of Generic Issue A-30, October 28,1991

UFSAR Section 8.3.2

DC Power Systems

NRC Information Notice 94-80Inadequate

DC Ground Detection in

DC DistributionSystems

UFSAR Section 9.3.4.3.3Reactor Makeup Control System

UFS [[]]

AR Section 3A.4.2.1Boration SystemCondition Reports2006-0025812006-006510

2006-006978

2006-007038Section 1R05: Fire ProtectionProceduresFRP-30.0,

SH Basement

SC-3.2.11, Immediate Action - Screenhouse Fire

SC-3.3.1, Immediate Fire Notification

SC-3.4.1, Fire Brigade Captain and Control Room Personnel ResponsibilitiesCondition Reports 2007-0003962007-000397DocumentsGinna Station Fire Protection ProgramSection 1R06: Flood Protection MeasuresDrawings33013-2681Sump Pumps Drains and Sewage Pumps33013-1912Condensate Demineralizer Regeneration Waste Handling

A-3AttachmentProceduresM95, Annual Inspection Maintenance and Operational Check of the Backflow Protection SystemCondition Reports2005-65602007-1070Section 1R07: Heat Exchanger PerformanceDocumentsDA-ME-97-016,

CCW and

RHR [[]]

HX Performance Evaluation, Rev. 0, 10/9/98

DA -ME-98-138,

EDG Lube Oil and Jacket Water

HX Plugging Limits and Thermal Performanceat Limiting

SW Flows, Rev. 1, 10/29/98

DA -ME-98-139,

EDG Lube Oil and Jacket Water

HX SW D/P Limits, Rev. 1, 7/9/99

DA -

ME -99-067,

SA [[]]

FW Pump Room Coolers Thermal Performance Evaluation, Rev. 0, 4/12/00

DA -

ME -2003-039,

CCS [[]]

HX A & B Thermal Performance Testing, performed 9/16/03, Rev. 0,3/9/04SWSROP, Service Water System Reliability Optimization Program, Rev. 7, 4/26/06

Repetitive Task Number P301717,

EAC [[01A - Clean, Inspect, Eddy Current tube side []]

CCWHX [[], 4/19/06Repetitive Task Number P301709,]]

CMP -10-03-

ESW08A/ESW09A - Clean Inspect, Eddy

Current tube side [EDG HXs], 3/3/06

Repetitive Task Number P401084, Open Inspect Clean

SA [[]]

FW Pump Room Cooler per

M-11.34.12 M-37.130 M-93, 3/10/06

CMP -10-04-

EAC 01A, Maintenance for

EAC [[01A []]

CCW [[]]

HX [[] performed 12/13/04, Rev. 3,12/15/97]]

CMP -10-03-ESW08A, Maintenance for

ESW [[08A []]

EDG [[]]

JW [[]]

HX [[] performed 10/21/05, Rev. 4,2/23/05CMP-10-03-ESW09A, Maintenance for]]

ESW [[09A []]

EDG [[]]

LO [[]]

HX] performed 10/21/05, Rev. 5,3/8/05Section 1R11: Licensed Operator RequalificationRequalification Program Procedures/Documents:OTG-2.0, Annual Examination Instructions, Revision19OTG-2.1, Oral/Walkthrough Examination Instructions, Revision 8

OTG-2.2, Simulator Examination Instructions, Revision 39

OTG-2.3, Job Performance Measure Instructions, Revision 16

OTG-2.4, Written Examination Instructions, Revision 14

OTG-2.5, Exam Failure Review Process, Revision13

OTG -2.6, Dynamic Simulator Examination Scenarios, Revision 15

OTG -2.8,

NRC Exam Security, Revision 13

OTG-10.0, License Activation, Revision3

TR -C.5.2, Licensed Operator Requalification Program, Revision 31

IP -

TQS-3, Operator and Fire Brigade Physicals, Revision 5

NTG -5, Exam Trouble Card System, Revision 4

OPS -

SHIFT ORG, Shift Organization, Revision 7

OPS -

SHIFT-SCHEDULE, Shift Scheduling and Watch Standing Requirements, Revision 4

2005/2006 Ginna

LO [[]]

RT Exam Sample Plan

Training Work Request 05-0483

A-4AttachmentSimulator-Related DocumentationSimulator Test Summaries for 2005 Model Upgrade Factory Acceptance Test, Site AcceptanceTest and 2006 Extended Power Uprate2006 Operating Limits Monitoring Test 14.4.1, Revision 10

2006 100% Steady State Accuracy Test 14.4.3.1, Revision 13

2006 NSSS -

BOP Energy Balance Test 14.4.4.1

List of Open Simulator Deficiency Reports

List of Simulator Deficiency Reports Closed 1/1/05 thru 12/31/06

2006 Sim Core Perf Tests 14.4.6.1 Revision 13 and 14.4.6.5 Revision 0

2006 Sim Malf Test 14.4.7.3.2, "CND-2, Main Condenser Tube Leak," Revision 8

2006 Sim Malf Test 14.4.7.4.2, "CVC-2, Letdown Line Leak Outside Cntmt," Revision 8

2006 Sim Malf Test 14.4.7.5.7, "EDS-07, Loss of Instrument Bus Supply," Revision 9

SDR -2004-130, Evaluate Concern -

BE 06 Response

SDR -2006-014,

PPCS Response to Bad Instrument Input

SDR-2006-021, Stm Line Break Inside Cntmt Does Not Have Restricted Flow

SDR -2006-026, Loss of Condensate Pumps Following a Trip

SDR -2006-032,

TDAFW Unexpected Trip During ECA-0.0

SDR -2006-037, Feedwater Pumps Trip Unexpectedly During

ATWS ActionsCondition ReportsCR-2007-0722CR-2007-0721CR-2007-0720CR-2007-0716CR-2007-0594CR-2007-0532CR-2006-7178CR-2006-7177

CR -2006-7087

CR -2006-6534CR-2006-6458CR-2006-6408

CR -2006-3637

CR -2006-3939CR-2006-3395CR-2006-3336

CR -2006-3168

CR -2006-3251CR-2006-2205CR-2006-1662

CR -2006-1500

CR -2006-1453CR-2006-1364CR-2006-0535

CR -2006-0467

CR -2006-2072CR-2005-6663CR-2005-2085

CR -2005-0756

CR-2006-0067Biennial Written Exams 2007Exams for Weeks One and Three

Reviewed Scenarios and

JPM [[s - 2007 Annual Operating ExamsExams for Weeks Two, Three, and FiveSection 1R12: Maintenance Rule ImplementationDrawings33013-1867Control Room Emergency Air Treatment System Cooling SystemCondition Report2007-000115Section 1R13: Maintenance Risk Assessments and Emergent Work EvaluationProcedures]]

PT-2.1S, A, B & C Safety Injection Pump Service Water Cooler Discharge Flow CheckRSSP-19, Diesel Generator "A" - Auto-Start Undervoltage Logic Test

S-3.2E, Placing in or Removing from Service Normal Letdown/Excess Letdown

AP -

FW. 1, Abnormal Main Feed Water Pump Flow or

NP [[]]

SH A-5AttachmentWork Orders20605735Replace

HC -135A, Letdown Pressure Hand ControllerCondition Reports2007-0001462007-0014882007-0014992007-0015002007-0015542007-0017842007-0024672007-002474DocumentsP&

ID 33013-2280 Gas Analyzer SkidUpdated Final Safety Analysis Report (UFSAR), Section 11.3.2.2.4, Gas AnalyzerUFSAR, Section 11.5.1, Process and Effluent Radiation Monitoring and Sampling SystemsDesign Bases

CHA -

EGSTRM Explosive Gas and Storage Tank Radioactivity Monitoring Program

EWR 4221, Revision 1, Safety Analysis Ginna Station O2/H2 Analyzer ReplacementSection 1R15: Operability EvaluationsCondition Reports2007-0008832007-000876

2007-000729

2007-002130

2007-002021.DocumentsFebruary 10, 1977 Letter From

RG&E to the

NRCJ une 8, 1977 Letter From

RG&E to the

NRC

Licensee Event Report 76-24 Section 1R17: Permanent Plant ModificationsDocumentsPCR 2006-0031Control Room and Simulator Furniture UpgradeCondition Reports 2007-0051292007-001540Section 1R19: Post Maintenance TestingProceduresPT-36Q-D, Standby Auxiliary Feedwater Pump D - QuarterlyPT-12.1, Emergency Diesel Generator "A"

PT -31, Charging Pump Inservice Test

CPI -

FT -464, Calibration of Steam Generator A Steam Flow Transmitter FT-464

PT -60.13A, Control Room Emergency Air Treatment System Heating and Cooling SystemPerformance Test - Train "A"

PT -17.1, Performance Test of Area Radiation Monitors and High Range Area RadiationMonitorsPT-2.10.5,

MS [[]]

IV Shutdown Exercising Requirements

PT-22.1,Equipment Hatch Door Seal Leakrate Test

A-6AttachmentWork Orders20504819Perform

PM Inspection on 52/

SAFWP1D, Standby Auxiliary FeedwaterPump 1D Breaker20700963"A" Charging Pump Belt Disintegrated During Operation. Replace Belt

20500843Replace the "A" Diesel Generator Fuel Oil Booster Pump as per

GMM -15-01-

KDG 01A/B Procedure20701655Swagelock Fitting on top of condensing Pot is Leaking.

FT -464Damaged When Root Valves Closed Prior to Transmitter IsolationDocuments

CME -50-02-52/SAFWPD, Rev 4, Westinghouse 480V Air Circuit Breaker Type

DB -50Standby Auxiliary Feedwater Pump 'D' Breaker

UFSAR 3.1.1.5.4, Reactivity Hold Down Capability

UFS [[]]

AR 3.1.2.4.4, General Design Criterion 33 - Reactor Coolant Makeup

Prompt Investigation Report for

CR 2007-002566Condition Reports2007-0012242007-002566Section 1R20: Refueling and Outage ActivitiesProcedures

PT -20, Infrared Thermography on Secondary Side Relief ValvesDocumentsGinna Technical Evaluation of March 16, 2007 Reactor TripPrompt Investigation Report for

CR 2007-002129, "B"

MSIV Failed closed at 100% Power Section 1R22: Surveillance TestingProceduresPT-12.2, Emergency Diesel Generator BPT-32, Reactor Trip Breaker Testing - Train B

PT -12.3, Security Emergency Diesel Test

PT -2.2Q, Residual Heat Removal System - QuarterlySection 1R23: Temporary Plant ModificationsProcedures

IP-DES-3, Temporary Modifications

Section

4OA 1: Performance Indicator VerificationDocuments

NEI 99-02, Nuclear Assessment Performance Indicator Guideline, Revision 4

Section 4OA2: Identification and Resolution of ProblemsProcedures-71.2Component Rework/Test and Relay Inspection Procedure

A-7AttachmentCondition Reports2006-0061322006-0067492006-0071412007-001032007-0002432007-002502007-0011292007-001130

2007-001668Section

4OA 3: Event Follow-upDocuments

PT -13.4.19Flood Valve Testing - Suppression System SO6 Air Handling Room AutoDelugeA-25.4Post-Trip Review Report

LER 05000244/2006007-01Main Steam Safety Valve Setpoint Exceedance Condition Reports 2007-0002832007-0002902007-001032007-0021262007-0021292007-0021342007-0021612007-002172

LIST [[]]

OF [[]]

ACRONYMSADAMSAgency-Wide Documents Access and Management SystemCAPcorrective action programCCWclosed cooling waterCFRCode of Federal RegulationCRcondition reportCREATScontrol room emergency air treatment systemCVCSchemical and volume control system DCdirect current

EDGemergency diesel generator

EHC electro-hydraulic control

EP [[]]

IPemergency plan implementing procedure

EROemergency response organization

GLgeneric letter

HXheat exchanger

IMCinspection manual chapter

INinformation notices

ISTin-service testing

JPM sjob performance measures

LO [[]]

CA loss-of-coolant accident

MS [[]]

IVmain steam isolation valve

MSR main steam reheater

MS [[]]

SVmain steam safety valve

NCV non-cited violation

NE [[]]

IN uclear Energy Institute

NR [[]]

CU.S. Nuclear Regulatory Commission

PA [[]]

RS Publicly Available Records

PCR plant change request

P& [[]]

IDpiping and instrument drawings

PI&R problem, identification and resolution

PS [[]]

IG pounds per square inch

RC [[]]

CArod cluster control assembly

RTPrated thermal power

A-8AttachmentSAFstandby auxiliary feedwaterSDPsignificance determination process

SHscreenhouse

SSCsystems, structures and components

TB turbine building

TDA [[]]

FWturbine driven auxiliary feedwater

TS technical specifications

UFSA [[]]

RU pdated Final Safety Analysis Report