ML091250233

IR 05000244-09-002; 01/01/2009 - 03/31/2009; R.E. Ginna Nuclear Power Plant (Ginna), Identification and Resolution of Problems, Followup of Events and Notices of Enforcement Discretion

ML091250233
Person / Time
Site:	Ginna  (DPR-018)
Issue date:	05/04/2009
From:	David Lew Division Reactor Projects I
To:	John Carlin Ginna
Lew D
References
EA-09-045 IR-09-002
Download: ML091250233 (36)

See also: IR 05000244/2009002

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD

KING OF PRUSSIA, PA 19406-1415

May 4, 2009

EA-09-045

Mr. John T. Carlin

Vice 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 INSPECTION

REPORT 05000244/2009002; PRELIMINARY WHITE FINDING

Dear Mr. Carlin:

On March 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection

at your R.E. Ginna Nuclear Power Plant. The enclosed integrated inspection report documents

the inspection results, which were discussed on April 16, 2009, 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 letter transmits one self-revealing finding that, using the reactor safety Significance

Determination Process (SDP), has preliminarily been determined to be White, a finding with low

to moderate safety significance. The finding is associated with inadequate implementation of

the preventive maintenance (PM) program for the turbine-driven auxiliary feedwater (TDAFW)

pump governor that led to a failure of the pump to operate properly during a December 2, 2008,

surveillance test. Following the test failure, Ginna replaced several components in the TDAFW

governor system, revised the TDAFW PM program, and successfully completed the surveillance

test. There is no immediate safety concern present due to this finding because the system is

now operable and the long term corrective actions are being implemented in Ginnas corrective

action program. The final resolution of this finding will be conveyed in separate

correspondence.

The finding is also an apparent violation of NRC requirements and is being considered for

escalated enforcement action in accordance with the enforcement policy, which can be found

on the NRCs Web site at http://www.nrc.gov/reading-rm/doc-collections/enforcement/.

In accordance with the NRC Inspection Manual Chapter (IMC) 0609, we intend to complete our

evaluation using the best available information and issue our final determination of safety

J. Carlin

2

significance within 90 days of the date of this letter. The significance determination process

encourages an open dialogue between the NRC staff and the licensee; however, the dialogue

should not impact the timeliness of the staffs final determination. Before we make a final

decision on this matter, we are providing you with an opportunity (1) to attend a Regulatory

Conference where you can present to the NRC your perspective on the facts and assumptions

the NRC used to arrive at the finding and assess its significance, or (2) submit your position on

the finding to the NRC in writing. If you request a Regulatory Conference, it should be held

within 30 days of the receipt of this letter and we encourage you to submit supporting

documentation at least one week prior to the conference in an effort to make the conference

more efficient and effective. If a Regulatory Conference is held, it will be open for public

observation. If you decide to submit only a written response, such submittal should be sent to

the NRC within 30 days of your receipt of this letter. If you decline to request a Regulatory

Conference or submit a written response, you relinquish your right to appeal the final SDP

determination, in that by not doing either you fail to meet the appeal requirements stated in the

Prerequisite and Limitation Sections of Attachment 2 of IMC 0609.

Please contact Glenn Dentel at 610-337-5233, and in writing, within 10 days from the issue date

of this letter to notify the NRC of your intentions. If we have not heard from you within 10 days,

we will continue with our significance determination and enforcement decision, and you will be

advised of the results of our deliberations on this matter.

Since the NRC has not made a final determination in this matter, no Notice of Violation is being

issued for this inspection finding at this time. In addition, please be advised that the number

and characterization of the apparent violation may change as a result of further NRC review.

In addition, the report documents one self-revealing finding of very low safety significance

(Green). The finding did not involve a violation of NRC requirements. If you disagree with the

characterization of any finding in this report, you should provide a response within 30 days of

the date of this inspection report, with the basis for your disagreement, to the Regional

Administrator, Region I, and the NRC Resident Inspector at R.E. Ginna Nuclear Power Plant.

The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and 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 the

NRCs document system (ADAMS). ADAMS is accessible from the NRC Web Site at

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/ Original Signed By;

David C. Lew, Director

Division of Reactor Projects

Docket No.: 50-244

License No.: DPR-18

J. Carlin

3

Enclosure: Inspection Report No. 05000244/2009002

w/ Attachment: Supplemental Information

cc w/encl:

M. J. Wallace, Vice - President, Constellation Energy

B. Barron, President, CEO & Chief Nuclear Officer, Constellation Energy Nuclear Group, LLC

P. Eddy, Electric Division, NYS Department of Public Service

C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law

C. Fleming, Esquire, Senior Counsel, Nuclear Generation, Constellation Nuclear Energy

Nuclear Group, LLC

T. Harding, Acting Director, Licensing, Constellation Energy Nuclear Group, LLC

A. Peterson,SLO Designee, New York State Energy Research and Development Authority

F, Murray, President & CEO, New York State Energy Research and Development Authority

G. Bastedo, Director, Wayne County Emergency Management Office

M. Meisenzahl, Administrator, Monroe County, Office of Emergency Management

T. Judson, Central New York Citizens Awareness Network

J. Carlin

2

significance within 90 days of the date of this letter. The significance determination process

encourages an open dialogue between the NRC staff and the licensee; however, the dialogue

should not impact the timeliness of the staffs final determination. Before we make a final

decision on this matter, we are providing you with an opportunity (1) to attend a Regulatory

Conference where you can present to the NRC your perspective on the facts and assumptions

the NRC used to arrive at the finding and assess its significance, or (2) submit your position on

the finding to the NRC in writing. If you request a Regulatory Conference, it should be held

within 30 days of the receipt of this letter and we encourage you to submit supporting

documentation at least one week prior to the conference in an effort to make the conference

more efficient and effective. If a Regulatory Conference is held, it will be open for public

observation. If you decide to submit only a written response, such submittal should be sent to

the NRC within 30 days of your receipt of this letter. If you decline to request a Regulatory

Conference or submit a written response, you relinquish your right to appeal the final SDP

determination, in that by not doing either you fail to meet the appeal requirements stated in the

Prerequisite and Limitation Sections of Attachment 2 of IMC 0609.

Please contact Glenn Dentel at 610-337-5233, and in writing, within 10 days from the issue date

of this letter to notify the NRC of your intentions. If we have not heard from you within 10 days,

we will continue with our significance determination and enforcement decision, and you will be

advised of the results of our deliberations on this matter.

Since the NRC has not made a final determination in this matter, no Notice of Violation is being

issued for this inspection finding at this time. In addition, please be advised that the number

and characterization of the apparent violation may change as a result of further NRC review.

In addition, the report documents one self-revealing finding of very low safety significance

(Green). The finding did not involve a violation of NRC requirements. If you disagree with the

characterization of any finding in this report, you should provide a response within 30 days of

the date of this inspection report, with the basis for your disagreement, to the Regional

Administrator, Region I, and the NRC Resident Inspector at R.E. Ginna Nuclear Power Plant.

The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and 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 the

NRCs document system (ADAMS). ADAMS is accessible from the NRC Web Site at

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/ Original Signed By:

David C. Lew, Director

Division of Reactor Projects

SUNSI Review Complete: gtd (Reviewers Initials) NAME: G:\\DRP\\BRANCH1\\Ginna\\Reports\\2009-002\\2009-002

Draft IR and Feedersrev 2.docAfter declaring this document An Official Agency Record it will be released to the

Public.

ML091250233

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OFFICE

RI/DRP

RI/DRP

RI/DRP

RI/DRS

RI/ORA

NAME

KKolaczyk/ksk

JHawkins/jrh

GDentel/gtd

WCook/wac

DHolody/djh

DATE

04/30/09

04/29/09

04/30/09

04/29/09

04/30/09

OFFICE

RI/DRP

NAME

DLew/dcl

DATE

05/04/09

J. Carlin

3

OFFICIAL RECORD COPY

Distribution w/encl:

S. Collins, RA

M. Dapas, DRA

D. Lew, DRP

J. Clifford, DRP

Stephen Campbell, RI OEDO

R. Nelson, NRR

D. V. Pickett, PM, NRR

B. Vaidya, PM, NRR

G. Dentel, DRP

N. Perry, DRP

J. Hawkins, DRP

K. Kolaczyk, DRP, SRI

M. Marshfield, DRP, RI

M. Rose, DRP, Resident OA

D. Bearde, DRP

Region I Docket Room (with concurrences)

ROPreports.Resource@nrc.gov

Email distribution to licensee

Enclosure

1

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No.:

50-244

License No.:

DPR-18

Report No.:

05000244/2009002

Licensee:

R.E. Ginna Nuclear Power Plant, LLC

Facility:

R.E. Ginna Nuclear Power Plant

Location:

Ontario, New York

Dates:

January 1, 2009 through March 31, 2009

Inspectors:

K, Kolaczyk, Senior Resident Inspector

L. Casey, Resident Inspector

M. Marshfield, Resident Inspector

W. Cook, Senior Reactor Analyst

D. Silk, Senior Operations Engineer

J. Hawkins, Project Engineer

S. Ibarrola, Reactor Engineer

Approved by:

Glenn T. Dentel, Chief

Projects Branch 1

Division of Reactor Projects

Enclosure

2

TABLE OF CONTENTS

SUMMARY OF FINDINGS ......................................................................................................... 3

REPORT DETAILS ..................................................................................................................... 5

1.

REACTOR SAFETY ........................................................................................................... 5

1R01 Adverse Weather Protection ................................................................................ 5

1R04 Equipment Alignment .......................................................................................... 5

1R05 Fire Protection .................................................................................................... 7

1R06 Flood Protection Measures ................................................................................. 7

1R11 Licensed Operator Requalification Program ........................................................ 7

1R12 Maintenance Effectiveness ................................................................................. 9

1R13 Maintenance Risk Assessments and Emergent Work Control .......................... 10

1R15 Operability Evaluations ..................................................................................... 10

1R18 Plant Modifications ........................................................................................... 11

1R19 Post-Maintenance Testing ................................................................................ 11

1R22 Surveillance Testing ......................................................................................... 12

1EP6 Drill Evaluation .................................................................................................. 13

4.

OTHER ACTIVITIES ......................................................................................................... 13

4OA1 Performance Indicator Verification ................................................................... 13

4OA2 Identification and Resolution of Problems ......................................................... 13

4OA3 Followup of Events and Notices of Enforcement Discretion .............................. 18

4OA5 Other Activities .................................................................................................. 20

4OA6 Meetings, Including Exit .................................................................................... 21

ATTACHMENT: SUPPLEMENTAL INFORMATION ................................................................ 21

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-10

Enclosure

3

SUMMARY OF FINDINGS

IR 05000244/2009002; 01/01/2009 - 03/31/2009; R.E. Ginna Nuclear Power Plant (Ginna),

Identification and Resolution of Problems, Followup of Events and Notices of Enforcement

Discretion.

The report covered a three-month period of inspection by resident inspectors and region-based

inspectors. One apparent violation (AV) with potential low to moderate safety significance

(Preliminary White) 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). The cross-cutting aspect for each finding

was determined using IMC 0305, Operating Reactor Assessment Program. Findings for which

the SDP does not apply may be Green or be assigned a severity level after NRC management

review. The NRCs program for overseeing the safe operation of commercial nuclear power

reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated

December 2006.

A.

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Preliminary White. The inspectors identified an AV of Technical Specification 5.4.1.a,

Procedures, for the failure of the licensee to implement an effective preventive

maintenance (PM) program for the turbine-driven auxiliary feedwater (TDAFW) pump

governor linkage. Specifically, procedure M-11.5C, AFW Pump Minor Mechanical

Inspection and Maintenance, Revision 29, which includes steps for cleaning and

lubricating the TDAFW pump governor linkages, was not properly implemented. The

cleaning and lubrication steps were inappropriately deleted during the work planning

process for the PM scheduled on the TDAFW system. As a result, the governor linkages

were not lubricated during the March 2008 maintenance period, which directly

contributed to the failure of the TDAFW pump as demonstrated by testing performed on

December 2, 2008. Ginnas planned corrective actions include increased frequency of

testing to validate the identified root cause and appropriate resolution, upgrades to the

maintenance procedure for disassembly and lubrication of bearing wear surfaces and

linkages, and guidance on the type of lubricant to use. In addition, corrective actions

include enhancements to the scope of minor maintenance requirements on the TDAFW

pump to ensure that the linkage cleaning and lubrication is not missed, and establishing

a 9-year periodicity to rebuild the governor and associated linkages.

The inspectors determined that this finding is more than minor because it is associated

with the procedure quality attribute of the Mitigating Systems Cornerstone and affects

the cornerstone objective to ensure the availability, reliability, and capability of systems

that respond to initiating events to prevent undesirable consequences. Specifically, the

failure to perform adequate maintenance resulted in the inoperability of the TDAFW

pump. This finding was assessed using IMC 0609 and preliminarily determined to be

White based on a Phase 3 analysis with a total (internal and external contributions)

calculated conditional core damage frequency (CCDF) of 8.8E-6. This finding has a

cross-cutting aspect in the area of human performance because Ginna did not establish

Enclosure

4

appropriate controls to assess how changes to the TDAFW PM program would impact

operation of the TDAFW system (H.3.b per IMC 0305). (Section 4OA2)

Green. A Green self-revealing finding was identified on February 5, 2009, when Ginna

failed to review applicable internal operating experience and implement compensatory

actions to minimize the consequences associated with replacement of the annunciator

cards, in accordance with CNG-OP-4.01-1000, Integrated Risk Management, Revision

00200. Specifically, CNG-OP-4.01-1000, requires work activities that are considered

medium risk to have contingency plans based in part on operating experience. As a

result, when the power supplies were inadvertently de-energized, restoration of the

alarm panels was delayed until recovery work instructions were prepared and

implemented. Ginnas corrective actions include adding a trouble shooting plan to work

packages for annunciators that depicts how to restore failed annunciators, revising CNG-

OP-4.01-1000, to incorporate a checklist of equipment important to the emergency plan

in the screening section of the risk process, and having an senior reactor operator

review the final weekly schedule for maintenance that could possibly impact equipment

used by the emergency plan.

This finding is more than minor because it is associated with the design control attribute

of the Mitigating Systems Cornerstone and affected the cornerstone objective of

ensuring the availability, reliability, and capability of systems that respond to initiating

events to prevent undesirable consequences. When the annunciator panels were de-

energized, the ability of operators to identify and respond to off-normal plant conditions

was degraded. Using Phase 1 of IMC 0609, Appendix A, the inspectors determined that

the finding was of low safety significance (Green), because the finding did not represent

a loss of system safety function; did not represent an actual loss of safety function of a

single train for greater than its Tech Spec allowed outage time; did not represent an

actual loss of safety function of one or more non-Tech Spec trains of equipment

designated as risk-significant per 10CFR50.65, for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and did not

screen as potentially risk significant due to a seismic, flooding, or severe weather

initiating event. This finding has a cross-cutting aspect in the area of human

performance because Ginna personnel did not appropriately plan work activities by

incorporating risk insights and the need for planned contingencies, compensatory

actions and abort criteria, which directly contributed to the loss of power to the control

board annunciator panels and declaration of an UE (H.3.a per IMC 0305). (Section

4OA3)

B.

Licensee-Identified Violations

None.

Attachment

5

REPORT DETAILS

Summary of Plant Status

R.E. Ginna Nuclear Power Plant (Ginna) began the inspection period operating at full-rated

thermal power and operated at full power for the entire period.

1.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01 - One sample)

a.

Inspection Scope

During the week of January 11, 2009, Ginna experienced unusually cold temperatures

with daytime high temperatures below 10 degrees. During this time, the inspectors

toured areas of the plant that contained equipment and systems that could be adversely

affected by cold temperatures. Areas of focus were the intake structure, auxiliary

building, the standby auxiliary feedwater (SAFW) pump room, and the A and B battery

and diesel generator rooms. During the tours, 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 performed field walkdowns of the

systems to verify that Ginna procedure O-22, Cold Weather Walkdown Procedure,

Revision 00500 was properly implemented. Documents reviewed for each inspection in

this report are listed in the Attachment.

b.

Findings

No findings of significance were identified.

1R04 Equipment Alignment (71111.04)

.1

Partial System Walkdown (71111.04Q - Three samples)

a.

Inspection Scope

The inspectors reviewed the alignment of system valves and electrical breakers to

ensure proper in-service or standby configurations as described in plant procedures,

piping and instrument drawings (P&ID), and the UFSAR. During the walkdown, the

inspectors evaluated the material condition and general housekeeping of the system and

adjacent spaces. The inspectors also verified that operators were following plant

technical specifications (TSs) and system operating procedures.

The following plant system alignments were reviewed:

6

Enclosure

On January 13, 2009, the inspectors performed a walkdown of the feed and

condensate water systems. These systems were selected based on recent industry

information and several feedwater related issues and concerns outlined in NRC

Operating Experience Smart Sample, FY 2009-02, Negative Trend and Recurring

Events Involving Feedwater Systems, Rev. 0. During this walkdown, valve

positions in major system flow paths were compared to the positions contained in

system drawings 33013-1252, Condensate, Rev. 23; 33013-1235, Condensate,

Rev. 20; 33013-1233, Condensate Low Pressure Feedwater Heaters, Rev. 29;

33013-1236, Feedwater, Sheet 1, Rev. 14; and 33013-1236, Feedwater, Sheet 2,

Rev. 13;

On February 3, 2009, the inspectors performed a walkdown of the D train of the

SAFW system while the A motor-driven AFW train was removed from service for

planned maintenance activities. During this walkdown, the inspectors compared

valve and breaker positions in major system flow paths to the positions contained in

system drawing 33013-1238, SAFW, Rev. 25, and procedure S-30.5, SAFW Pump

Valve and Breaker Position Verification, Rev. 34; and

On March 19, 2009, the inspectors performed a walkdown of the B diesel generator

and associated support systems while a new level indicating system was being

installed on the A diesel generator fuel oil storage tank. During this walkdown, the

inspectors compared valve and breaker positions to the positions contained in

system drawing 33013-1239, Diesel Generator B, Rev. 21.

b.

Findings

No findings of significance were identified.

.2

Complete Walkdown (71111.04S - One sample)

a.

Inspection Scope

The inspectors performed a detailed walkdown of the component cooling water (CCW)

system. CCW was chosen because of its risk significant function to provide cooling for

the residual heat removal (RHR) heat exchangers (HXs) and emergency core cooling

system pumps. Other functions of CCW include providing cooling to the reactor coolant

pumps, reactor support cooling pads, excess letdown HX, and the non-regenerative HX.

The inspectors verified proper system alignment as specified by TSs, UFSAR, P&IDs,

and plant procedures. Inspectors reviewed documentation associated with open

maintenance requests and items tracked by plant engineering to assess their collective

impact on system operation. In addition, the inspectors utilized the corrective action

database to verify that any equipment alignment problems were being identified and

appropriately resolved.

b.

Findings

No findings of significance were identified.

7

Enclosure

1R05 Fire Protection (71111.05)

Quarterly Inspection (71111.05Q - Five samples)

a.

Inspection Scope

The inspectors performed walkdowns of 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 inspected against Ginnas licensing basis and 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:

Technical Support Center (Fire Zone TSC-1S);

Auxiliary Building Operating Floor (Fire Zone ABO);

Cable Tunnel (Fire Area CT);

Relay Room (Fire Zone RR); and

SAFW Pump Building (Fire Area SAF).

b.

Findings

No findings of significance were identified.

1R06 Flood Protection Measures (71111.06 - One sample)

a.

Inspection Scope

The inspectors walked down the auxiliary building basement to verify Ginna had

implemented appropriate measures to reduce the possibility that the area could be

damaged by internal flooding. To perform this evaluation, the inspectors reviewed the

UFSAR, integrated plant safety assessment, condition reports (CRs), plant change

records (PCRs), the site repetitive task database, and various flooding analysis for

equipment located in the area of concern. During the field walkdown, to the extent

practicable, the condition of flood mitigation equipment in this area was examined by the

inspectors.

b.

Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program (71111.11)

.1

Resident Inspector Quarterly Review (71111.11Q - One sample)

a.

Inspection Scope

On January 21, 2009, the inspectors observed a licensed operator simulator scenario,

8

Enclosure

ES1213-05, Small Break Loss of Coolant Accident, Revision 9. 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, Revision 43.

b.

Findings

No findings of significance were identified.

.2

Biennial Review (71111.11B - One sample)

a.

Inspection Scope

The following inspection activities were performed using NUREG-1021, AOperator

Licensing Examination Standards for Power Reactors, Revision 9, Inspection Procedure

Attachment 71111.11, Licensed Operator Requalification Program, NRC Manual

Chapter 0609, Appendix I, Operator Requalification Human Performance Significance

Determination Process, and 10 CFR Part 55.

The inspectors reviewed documentation of operating history since the last requalification

program inspection. The inspectors also discussed facility operating events with the

resident staff. Documents reviewed included NRC inspection reports, licensee event

reports, Ginnas corrective action program (CAP), and the most recent NRC plant issues

matrix. The inspectors also reviewed specific events from Ginnas CAP that involved

human performance issues for licensed operators to ensure that operational events were

not indicative of possible training deficiencies.

The operating and written examinations for the week of January 12, 2009, were

reviewed for quality, performance, and excessive overlap.

On February 19, 2009, the results of the annual operating tests and the written exam for

2009 were reviewed to determine if pass fail rates were consistent with the guidance of

NUREG-1021 and NRC Manual Chapter 0609, Appendix I. The inspectors verified that:

$

Crew pass rates were greater than 80%. (Pass rate was 85.7%);

$

Individual pass rates on the written exam were greater than 80%. (Pass rate was

96.8%);

$

Individual pass rates on the job performance measures of the operating exam

were greater than 80%. (Pass rate was 96.8%); and

$

More than 75% of the individuals passed all portions of the exam. (93.5% of the

individuals passed all portions of the exam).

Observations were made of the dynamic simulator exams and job performance

measures (JPMs) administered during the week of January 12, 2009. These

observations included facility evaluations of crew and individual performance during the

dynamic simulator exams and individual performance of six JPMs.

9

Enclosure

The remediation plans for a crew/individual=s failure and a written exam failure were

reviewed to assess the effectiveness of the remedial training.

Four license reactivations were reviewed to ensure that license conditions and

applicable program requirements were met.

Simulator performance and fidelity were reviewed for conformance to the reference plant

control room. Selected simulator deficiency reports were reviewed to assess licensee

prioritization and timeliness of resolution. Simulator testing records were reviewed to

verify that scheduled tests were performed.

A sample of records for requalification training attendance, program feedback, reporting,

and 10 operator medical reports were reviewed for compliance with license conditions,

including NRC regulations.

b.

Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q - Two samples)

a.

Inspection Scope

The inspectors evaluated work practices and follow-up corrective actions for selected

systems, structures, and components (SSCs) for maintenance effectiveness. The

inspectors reviewed the performance history of those SSCs and assessed extent-of-

condition determinations for those issues with potential common cause or generic

implications to evaluate the adequacy of 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 procedures and the requirements of 10 CFR Part 50.65, Requirements

for Monitoring the Effectiveness of Maintenance. In addition, the inspectors reviewed

selected SSC classifications, performance criteria and goals, and corrective actions that

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

The following issues were reviewed:

Control Room Emergency Air Treatment System (CREATS) train B breaker failure

(CR-2008-009624).

Failure of main steam atmospheric relief valve (ARV) B (AOV-3410) to close (CR-

2009-001218).

b.

Findings

No findings of significance were identified.

10

Enclosure

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - Four samples)

a.

Inspection Scope

The inspectors evaluated the effectiveness of Ginnas maintenance risk assessments

required by 10 CFR Part 50.65(a)(4). The inspectors discussed with control room

operators and scheduling department personnel required actions regarding the use of

Ginnas online risk monitoring software. The inspectors reviewed equipment tracking

documentation and daily work schedules, and performed plant tours to verify that actual

plant configuration matched the assessed configuration. Additionally, the inspectors

verified that risk management actions, for both planned and emergent work, were

consistent with those described in CNG-OP-4.01-1000, Integrated Risk Management,

Revision 00100.

Risk assessments for the following out-of-service SSCs were reviewed:

Planned monthly surveillance testing on the B emergency diesel generator (EDG)

during a cold weather condition (January 14, 2009);

Emergent failure of main control room annunciator panels during maintenance

activities (February 5, 2009);

The week of March 8, 2009, included planned maintenance for the B train of the

RHR system, testing of the B diesel generator, and B train reactor trip breaker

testing; and

Planned removal of concrete structures adjacent to the buried auxiliary building

service water (SW) supply and return piping (March 25 to 31, 2009).

b.

Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15 - Five samples)

a.

Inspection Scope

The inspectors reviewed operability evaluations and/or CRs in order to verify that the

identified conditions did not adversely affect safety system operability or plant safety.

The evaluations were reviewed using criteria specified in NRC Regulatory Issue

Summary 2005-20, Revision to Guidance formerly contained in NRC Generic Letter 91-

18, Information to Licensees Regarding Two NRC Inspection Manual Sections on

Resolution of Degraded and Nonconforming Conditions and on Operability and

Inspection Manual Part 9900, Operability Determinations and Functionality

Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to

Quality or Safety. In addition, where a component was inoperable, the inspectors

verified the TS limiting condition for operation implications were properly addressed.

The inspectors performed field walkdowns, interviewed personnel, and reviewed the

following items:

CR 2009-0242, EDG Day Tank Level Set Points;

11

Enclosure

CR 2009-0437, Potential Error in Safety Injection (SI) Accumulator Low Pressure

Surveillance Limit;

CR 2009-0738, Motor-Operated Valve (MOV) 4007 Design Analysis Does Not

Account For Worst Case Operational Scenario;

CR 2009-1305, EDG Jacket Water HX Leak; and

CR 2009-0903, Slightly Lowering Oil Level On RCP 1A Bearing.

b.

Findings

No findings of significance were identified.

1R18 Plant Modifications (71111.18 - One sample)

Permanent Modification

a.

Inspection Scope

The inspectors reviewed PCR 2008-0034, Installation of Rupture Disks Upstream of the

SW Thermal Relief Valves, Revision 0. The inspectors reviewed the PCR to ensure that

the installation of the rupture disk would not adversely affect pressure relief capability

and that the material classification and functional properties were consistent with the

design basis and were compatible with installed SSCs. The inspectors verified that

affected procedures, drawings, and analysis were identified and that necessary changes

were captured in the PCR.

b.

Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19 - Five samples)

a.

Inspection Scope

The inspectors observed portions of post-maintenance testing (PMT) activities in the

field to determine whether the tests were performed in accordance with approved

procedures. The inspectors assessed each tests adequacy by comparing the test

methodology to the scope of maintenance performed. In addition, the inspectors

evaluated the test acceptance criteria to verify that the tested components satisfied the

applicable design, licensing bases and TS requirements. The inspectors reviewed the

recorded test data to determine whether the acceptance criteria were satisfied.

The following PMT activities were reviewed:

STP-O-2.2QB, RHR Pump B Inservice Test, Rev. 00101, to test the B RHR train

after installation of a relief valve modification performed under work order (WO)

20805574 (January 5, 2009);

12

Enclosure

GME-45-99-01, Electric Motor Inspection and Maintenance, Rev. 02101, to retest a

component cooling water pump breaker under WO 20807112, Perform Electrical

Tests on Breaker MO/CF1B (January 27, 2009);

STP-O-12.2, EDG B, Rev. 00301, to test the B EDG after jacket water HX

maintenance due to tube leaks under WO 20900978, Open, Inspect, Repair

ESW08B (March 2, 2009);

STP-O-12.1, EDG A, Rev. 00401, to test the A EDG after fuel oil day tank check

valve work under WO 20800872, Perform Major Inspection of CV-5960A

(March 3, 2009); and

STP-O-2.2QB, RHR Pump B Inservice Test, Rev. 00101, to test the B RHR train

after pump and valve maintenance under WOs 20805650, 20805651, 20805665,

and 20900937, B RHR Functional Equipment Group Maintenance Window

(March 9, 2009).

b.

Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - Six samples)

a.

Inspection Scope

The inspectors observed the performance and/or reviewed test data for the following

surveillance tests that are associated with selected risk-significant SSCs to verify that

TSs were followed and that acceptance criteria were properly specified. The inspectors

also verified that proper test conditions were established as specified in the procedures,

no equipment preconditioning activities occurred, and acceptance criteria were met.

STP-O-22.2, Local Leak Rate Test of Personnel Hatch Door Seal, Rev. 00003

(January 26, 2009) (IST LLRT)

STP-O-12.2, EDG B, Rev. 00301 (February 11, 2009) (IST)

PT-16Q-T, AFW Turbine Pump - Quarterly, Rev. 05801 (February 12, 2009) (IST)

PT-36Q-C, SAFW Pump C - Quarterly, Rev. 05700 (February 18, 2009) (IST)

STP-O-2.8Q, CCW Pump - Quarterly Test, Rev. 00002 (March 14, 2009) (IST)

STP-O-16Q-B, AFW Pump B - Quarterly, Rev. 00300 (March 26, 2009) (IST)

b.

Findings

No findings of significance were identified.

13

Enclosure

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation (71114.06 - One sample)

a.

Inspection Scope

On January 21, 2009, the inspectors observed a licensed operator simulator scenario,

ES1213-05, Small Break Loss of Coolant Accident, Revision 9, which included a

limited test of Ginnas emergency response plan. The inspectors verified that

emergency classification declarations and notifications were completed in accordance

with 10 CFR Part 50.72, 10 CFR Part 50 Appendix E, and the site emergency plan

implementing procedures.

b.

Findings

No findings of significance were identified.

4.

OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151)

Cornerstone: Initiating Events

a.

Inspection Scope (71151 - Three samples)

Using the criteria specified in Nuclear Energy Institute (NEI) 99-02, Regulatory

Assessment Performance Indicator (PI) Guideline, Revision 5, the inspectors verified

the completeness and accuracy of the PI data for calendar year 2008 for unplanned

scrams per 7,000 critical hours, unplanned power changes per 7,000 critical hours, and

unplanned scrams with complications. To verify the accuracy of the data, the inspectors

reviewed monthly operating reports, NRC inspection reports, and Ginna event reports

issued during 2008.

b.

Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152 - One sample)

.1

Continuous Review of Items Entered into the Corrective Action Program

a.

Inspection Scope

As specified 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

Ginnas CAP. This review was accomplished by reviewing electronic copies of CRs,

periodic attendance at daily screening meetings, and accessing Ginnas computerized

14

Enclosure

database.

b.

Findings

No findings of significance were identified.

.2

Annual Sample - TDAFW Pump Surveillance Test Failure (71152 - One sample)

a.

Inspection Scope

The inspectors reviewed the troubleshooting activities implemented by Ginna personnel

to identify and correct the cause for a failed surveillance test performed on the TDAFW

pump in December 2008. The review included examining components in the plant,

interviewing personnel, and examining a Ginna root-cause report.

b.

Findings and Observations

Introduction: The inspectors identified an apparent violation (AV) of TS 5.4.1.a,

Procedures, for a failure of Ginna to implement an effective PM program for the

TDAFW pump governor linkages in accordance with Ginna procedures. Specifically,

procedure M-11.5C, AFW Pump Minor Mechanical Inspection and Maintenance,

Revision 29, which includes steps for cleaning and lubricating the TDAFW pump

governor linkages was not implemented. The cleaning and lubrication steps were

inappropriately deleted during the work planning process for the PM scheduled on the

TDAFW system. As a result, the governor linkages were not lubricated during the March

2008 maintenance period, which directly contributed in the failure of the TDAFW pump

during testing performed on December 2, 2008.

Description: On December 2, 2008, Ginna performed a test of the TDAFW pump

system in accordance with procedure PT-16Q-T, AFW Turbine PumpQuarterly,

Revision 05801. During this test, the pump did not develop the minimum acceptable

discharge flow and pressure. The pump was declared inoperable and an incident

response team was formed to investigate the cause of the test failure. Oil samples from

the governor control system were taken for analysis, and the vendor was contacted.

Troubleshooting eventually revealed that the governor linkage stuck preventing the

pump from developing the required pump head and flow to satisfy the test.

Initial troubleshooting involved removal of a pin from the governor linkage and

verification of adequate freedom of movement of the relay valve, the servo arm, and the

control valve arm. The inlet steam check valves were also verified to be functional. The

quarterly test was re-performed after this initial troubleshooting and all TDAFW pump

performance parameters were satisfied. Oil sample results subsequently became

available and based on a higher than expected particulate count (although still within

specification), Ginna replaced the governor. Upon retesting the system, after the

governor was replaced, the speed of the turbine was unable to be adjusted and a

linkage pin was noted to be stuck halfway up the yoke arm at the bottom of the servo

arm. The linkage was then disassembled, cleaned, and lubricated with a dry lubricant

suitable for a high temperature environment. A more comprehensive surveillance test

involving full flow to the steam generators was then performed, the governor was

15

Enclosure

adjusted, and the TDAFW pump was restored to an operable condition. The

troubleshooting and maintenance resulted in slightly less than 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> of unscheduled

unavailability time for the TDAFW pump.

Ginnas root cause team evaluated the TDAFW pump failure and determined that during

the last scheduled maintenance window for the TDAFW pump in March 2008, the

governor linkages were not lubricated because steps in procedure M-11.5C that

lubricate the linkages, were deleted during the maintenance planning process. The lack

of proper lubrication in the governor linkage assembly caused the linkage to bind during

the December 2008 surveillance testing. The Ginna team identified the root cause of the

TDAFW pump failure to be inadequate managerial controls for the level of detail

described in the preventative maintenance scope, as described in the maintenance

repetitive task description. Additionally, Ginna determined that no specific barrier

existed to ensure that the requirements of the repetitive task were met, and that no

linkage lubrication standard existed to ensure that the proper type of lubrication was

used and that the proper scope of cleaning was performed.

The inspectors reviewed the root cause evaluation and associated corrective actions.

Planned corrective actions include increased frequency of testing to validate the

identified root cause and appropriate resolution, upgrades to the maintenance procedure

for disassembly and lubrication of bearing wear surfaces and linkages, and guidance on

the type of lubricant to use. In addition, corrective actions include enhancements to the

scope of minor maintenance requirements on the TDAFW pump to ensure that the

linkage cleaning and lubrication is not missed, and establishing a 9-year periodicity to

rebuild the governor and associated linkages. The 9-year rebuild is within the vendors

recommended 10-year service life for the TDAFW pump governor.

Analysis: The performance deficiency associated with this event is that Ginna did not

implement an adequate PM program for the TDAFW pump governor linkages.

Specifically, during planning for March 2008 PM activities on the TDAFW pump, steps

for cleaning and lubricating the governor linkage were deleted from procedure, M-11.5C.

As a result, during a quarterly surveillance test on December 2, 2008, the governor

control linkage, which had not been properly lubricated in March 2008, did not operate

properly which caused the pump to fail to develop the required discharge flow and

pressure.

The inspectors determined that this finding is more than minor because it is associated

with the procedure quality attribute of the Mitigating Systems Cornerstone and affects

the cornerstone objective to ensure the availability, reliability, and capability of systems

that respond to initiating events to prevent undesirable consequences. Specifically, the

failure to conduct adequate maintenance resulted in inoperability of the TDAFW pump.

In accordance with IMC 0609, Significance Determination Process, Phase 1

worksheets, a Phase 2 risk analysis was required because the finding represents an

actual loss of safety function of a single train for greater than the TS allowed outage time

of 7 days.

The Phase 2 risk evaluation was performed in accordance with IMC 0609, Appendix A,

Attachment 1, User Guidance for Significance Determination of Reactor Inspection

Findings for At-Power Situations. Because the precise time is unknown for the

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Enclosure

inception of TDAFW pump inoperability, an exposure time of one-half of the time period

(t/2) between discovery (December 2, 2008) to the last successfully completed quarterly

surveillance test (September 3, 2008) was used. This t/2 exposure time equals 45 days.

Using Ginnas Phase 2 SDP notebook, pre-solved worksheets, and an initiating event

likelihood of 1 year (>30-days exposure time), the inspector identified that this finding is

of potentially substantial safety significance (Yellow). The dominant sequence identified

in the Phase 2 notebook involves a loss of offsite power (LOOP), failure of both

EDGs, and the subsequent loss of the TDAFW pump, with the failure of operators to

restore offsite power within 1 hour: LOOP (2) + EAC (3) + TDAFW (0) + REC1 (0) = 5

(Yellow). In recognition that the Phase 2 notebook typically yields a conservative result,

a NRC Region I Senior Reactor Analyst (SRA) performed a Phase 3 risk assessment of

this finding.

The SRA used Ginnas Standardized Plant Analysis Risk (SPAR) model, Revision 3.45,

dated June 2008, and graphical evaluation module, in conjunction with the System

Analysis Programs for Hands-On Integrated Reliability Evaluations, Version 7, to

estimate the internal risk contribution of the Phase 3 risk assessment. The following

assumptions were used for this assessment:

To closely approximate the type of failure exhibited by the TDAFW pump, the SRA

used the TDAFW pump failure-to-run event <AFW-TDP-FR-TDP> and changed its

failure probability to 1.0, representing a 100 percent failure-to-run condition;

The exposure time for this condition was 1,125 hours0.00145 days <br />0.0347 hours <br />2.066799e-4 weeks <br />4.75625e-5 months <br /> (45 days, plus 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> of

unavailability during troubleshooting and repair);

Based upon the nature of the failure, no operator recovery credit was provided;

All remaining events were left at their nominal failure probabilities; and

Cut-set probability calculation truncation was set at 1E-13.

Based upon the above assumptions, the SPAR model internal contribution to conditional

core damage probability (CCDP) was calculated at 4.8E-6. The dominant internal event

sequences involved a loss of offsite power event with subsequent failure of one or both

EDGs (station blackout event) and/or the failure of a motor-driven AFW train. These

Phase 3 SPAR model results correlate well to the Phase 2 SDP notebook dominant core

damage sequences.

The SRA used Ginnas external risk assessment to quantify the external risk contribution

for this condition. Seismic event likelihood is very low and qualitatively determined to not

be a significant contributor to external event risk. Ginnas approved Probabilistic Risk

Analysis Evaluation Request No. G1-2009-002, dated February 27, 2009, identified the

external (fire) risk contribution associated with the failure of the TDAFW pump to be

3.3E-6. The risk contribution associated with flooding events was calculated to be 7.4E-

7. These delta CCDP values were based upon a 45-day exposure period. The most

significant fire-initiated core damage sequences involved a spectrum of control room

fires (with automatic and manual suppression failures) with subsequent failure of the

TDAFW pump, and the failure of operators to align the C SAFW pump for decay heat

removal via the steam generators. In addition, a relay room fire (with automatic and

manual suppression failures) with subsequent failure of the TDAFW pump, and failure of

operators to align the C SAFW pump, were identified as significant core damage

sequences. The most significant flooding core damage sequences quantified by Ginna

17

Enclosure

involved a large SW system line break/rupture in the auxiliary building. The SW system

supplies the component cooling water (CCW) system. Including the loss of CCW, as a

result of the SW line break, the flooding would cause the subsequent loss of charging

system (located in the basement elevation of the auxiliary building) and consequential

reactor coolant pump seal failure (small break loss of coolant accident).

The calculated total risk significance of this finding is based upon the summation of

internal and external risk contributions [delta CCDP internal + delta CCDP external (fires

and floods) = delta CCDP total]. 4.8E-6 + 3.3E-6 + 7.4E-7 = 8.8E-6 delta CCDP.

Annualized, this value of 8.8E-6 delta core damage frequency (CDF) represents a low to

moderate safety significance or White finding.

The Ginna containment is classified as a pressurized water reactor large dry

containment design. Based upon the dominant sequences involving loss of offsite

power and station blackout initiating events, per IMC 0609, Appendix H, Table 5.2,

Phase 2 Assessment FactorsType A Findings at Full Power, the failure of the

TDAFW pump does not represent a significant challenge to containment integrity early in

the postulated core damage sequences. Consequently, this finding does not screen as

a significant large early release contributor because the close-in populations can be

effectively evacuated far in advance of any postulated release due to core damage.

Accordingly, the risk significance of this finding is associated with the delta CDF value,

per IMC 0609, Appendix H, Figure 5.1, and not delta large early release frequency.

This finding has a cross-cutting aspect in the area of human performance because

Ginna did not establish appropriate controls to assess how changes to the TDAFW PM

program would impact operation of the TDAFW system (H.3.b per IMC 0305).

Enforcement: TS 5.4.1.a, Procedures, requires, in part, that the applicable procedures

recommended in Appendix A of Regulatory Guide (RG) 1.33, Quality Assurance

Program Requirements (Operations), shall be established, implemented and

maintained. RG 1.33, Appendix A, Section 9 (b), states, "PM schedules should be

developed to specify lubrication schedules, inspection of equipment, replacement of

such items as filters and strainers, and inspection or replacement of parts that have a

specific lifetime such as wear rings. Ginna procedure M-11.5C, Auxiliary Feedwater

Pump Minor Mechanical Inspection and Maintenance, Rev. 29, which is an 18-month

maintenance requirement for the TDAFW pump, contains steps which would have

properly conducted cleaning and lubrication maintenance on the governor linkage.

Contrary to the above, in March 2008, while performing PM on the TDAFW pump, Ginna

technicians used a procedure that did not implement the correct lubrication schedules.

Specifically, procedure M-11.5C, AFW Pump Minor Mechanical Inspection and

Maintenance, had steps for cleaning and lubricating the TDAFW pump governor

linkages that were deleted during the maintenance work planning. The lack of

lubrication led to the operational failure of the TDAFW pump as demonstrated by testing

on December 2, 2008. This issue was entered into Ginnas CAP as CR 2008-9911.

Pending final determination of significance, this finding is identified as an AV. (AV

05000244/2009002-01: Failure to Properly Lubricate Governor Linkage)

18

Enclosure

4OA3 Followup of Events and Notices of Enforcement Discretion (71153 - One sample)

Unusual Event Declaration for Loss of Four Annunciator Panels

a.

Inspection Scope

On February 5, 2009, at 1:58 p.m., during a planned maintenance activity on the MCB

annunciator system, Ginna experienced a failure of MCB annunciator panels E, F, G,

and H. At the time of the event, instrumentation and control (I&C) technicians were

replacing an annunciator card in control room panel H. In accordance with the Ginna

emergency plan, control room operators declared an Unusual Event (UE) at 2:13 p.m. in

accordance with emergency action level 7.3.1, Unplanned Loss of Annunciators or

Indications on any Control Room Panels for Greater Than 15 minutes. Subsequent

troubleshooting activities by Ginna personnel determined that the most likely cause of

the failure was an electrical spike, created by the annunciator card replacement activity

that caused the annunciator panel power supplies to down power into a preprogrammed

quiescent mode, which de-energized the annunciator panels. After Ginna verified that

the annunciator power supplies had not been damaged by the electrical spike, the power

supplies were reenergized to their normal full rated output level and the annunciator

panels were tested. Ginna terminated the UE at 4:35 a.m. on February 6, 2009.

The resident inspectors responded to the control room and technical support center to

evaluate the initial actions taken by operators in response to the loss of the annunciator

panels and to observe troubleshooting activities. Inspector activities included verifying

Ginna operators were adhering to the applicable emergency response procedures and

that troubleshooting activities were performed in a controlled manner. While the

annunciator panels were not functioning, additional operators were stationed in the

control room to monitor plant conditions using alternate systems such as the plant

process computer. The inspectors verified that appropriate compensatory measures

were in place to monitor plant parameters in the control room and the plant. During the

event, the inspectors performed tours to verify that the plant was maintained in a stable

condition and actions were in place to minimize the possibility of a plant transient.

Following the event, the inspectors interviewed Ginna I&C technicians who were

involved in the maintenance activity, operations personnel who were on shift during the

event, and reviewed the annunciator card replacement work instruction package.

b.

Findings

Introduction: A Green self-revealing finding was identified on February 5, 2009, when

Ginna failed to review applicable internal operating experience and implement

compensatory actions to minimize the consequences associated with replacement of the

annunciator cards, in accordance with CNG-OP-4.01-1000, Integrated Risk

Management. Due to this failure, Ginna I&C technicians inadvertently de-energized

main control board annunciator panels E, F, G, and H, which resulted in the subsequent

declaration of an UE.

Description: The Ginna control room operating board has three main control room

sections. Above each section are four annunciator panels that are powered by individual

19

Enclosure

power supplies. Each panel contains electronic card modules that inform operators of

potential off-normal plant conditions by generating a warning light and audible alarm. On

July 4, 2007, Ginna declared an UE when an age-related annunciator card failure

rendered several annunciator panels inoperable. To reduce the possibility of a

subsequent age-related card failure, Ginna began to replace the annunciator cards, the

majority of which had been in service since original plant construction, with reengineered

cards that were not susceptible to a similar age-related failure mechanism. At the time of

the February 5, 2009, event, Ginna I&C personnel had replaced all but 11 of the 300

control room annunciator cards.

The inspectors noted that the potential for the annunciator panel power supplies to down

power into a safe mode in the event of an electrical power spike was a known

vulnerability that was documented in a Ginna mechanical maintenance procedure.

Specifically, Ginna procedure M-94, Repair of RIS Alarm Panels in MCB, contained a

caution that stated, Electrical noise or excessive ripple on annunciator power supply

can cause converter lock-up, resulting in loss of an annunciator panel. Despite this

potential, the applicable work instructions for the card replacement activity did not have

adequate instructions to minimize the potential for this event to occur or sufficient

instructions to recover from this event if the power supplies were inadvertently de-

energized. This was contrary to the requirements outlined in Ginna procedure CNG-OP-

4.01-1000, Integrated Risk Management, which requires work activities that are

considered medium risk, which the card replacement activity was classified, to have

contingency plans to be based, in part, on operating experience. As a result, when the

power supplies were inadvertently de-energized, restoration of the alarm panels was

delayed until recovery work instructions were prepared and implemented.

Ginnas corrective actions include adding a trouble shooting plan to work packages for

annunciators that depicts how to restore failed annunciators, revising CNG-OP-4.01-

1000, Integrated Risk Management, to incorporate a checklist of equipment important

to the emergency plan in the screening section of the risk process, and having an senior

reactor operator review the final weekly schedule for maintenance that could possibly

impact equipment used by the emergency plan. In addition, corrective actions include

revising M-94, Repair of RIS Alarm Panels in Main Control Board (MCB), to provide

additional guidance on potential failure modes and require additional operations

compensatory measures and potential emergency action level (EAL) risk mitigation

during repair activities on the annunciators.

Analysis: The performance deficiency associated with this self-revealing finding involved

a failure of Ginna to review applicable internal operating experience and implement

compensatory actions to minimize the consequences associated with replacement of the

annunciator cards. Specifically, the work package that was being used by Ginna to

replace the annunciator cards, did not have instructions in place to mitigate a known

vulnerability concerning the annunciator panel power suppliesthe potential of the

supplies to de-energize in the event of a power spike. As a result, the annunciator

panels were inadvertently de-energized during the maintenance activity, and the panels

remained de-energized for over 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.

This finding is more than minor because it is associated with the design control attribute

of the Mitigating Systems Cornerstone and affected the cornerstone objective of

20

Enclosure

ensuring the availability, reliability, and capability of systems that respond to initiating

events to prevent undesirable consequences. When the annunciator panels were de-

energized, the ability of operators to identify and respond to off-normal plant conditions

was degraded. Using Phase 1 of IMC 0609, Appendix A, the inspectors determined that

the finding was of low safety significance (Green), because the finding did not represent

a loss of system safety function; did not represent an actual loss of safety function of a

single train for greater than its Tech Spec allowed outage time; did not represent an

actual loss of safety function of one or more non-Tech Spec trains of equipment

designated as risk-significant per 10CFR50.65, for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and did not

screen as potentially risk significant due to a seismic, flooding, or severe weather

initiating event.

This finding has a cross-cutting aspect in the area of human performance because

Ginna personnel did not appropriately plan work activities by incorporating risk insights

and the need for planned contingencies compensatory actions and abort criteria, which

directly contributed to the loss of power to the control board annunciator panels and

declaration of an UE (H.3.a per IMC 0305).

Enforcement: Enforcement action does not apply because the performance deficiency

did not involve a violation of a regulatory requirement and the control room annunciator

system is not a safety-related system. Additionally, the annunciator panel system failure

did not adversely impact safety-related systems. (FIN 05000244/2009002-02,

Inadequate Risk Management Results in Loss of Normal Control Room

Annunciators)

4OA5 Other Activities

Quarterly Resident Inspector Observations of Security Personnel and Activities

a.

Inspection Scope

During the inspection period, the inspectors conducted observations of security force

personnel and activities to ensure that the activities were consistent with Ginnas

security procedures and regulatory requirements relating to nuclear plant security.

These observations took place during both normal and off-normal plant working hours.

These quarterly resident inspector observations of security force personnel and activities

did not constitute any additional inspection samples. Rather, they were considered an

integral part of the inspectors normal plant status review and inspection activities.

b.

Findings

No findings of significance were identified.

21

Enclosure

4OA6 Meetings, Including Exit

.1

Annual Assessment Meeting Summary

On March 24, 2009, the Division of Reactors Projects Branch 1 Chief met with Ginnas

senior management to discuss the annual assessment letter, including the NRCs

assessment of Ginnas performance, and the NRCs inspection schedule.

.2

Exit Meeting Summary

On April 16, 2009, the resident inspectors presented the inspection results to

Mr. John Carlin and other members of his staff, who acknowledged the findings. The

inspectors verified that none of the material examined during the inspection is

considered proprietary in nature.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Attachment

A-1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

J. Carlin

Vice President, Ginna

D. Dean

Assistant Operations Manager (Shift)

M. Giacini

Scheduling Manager

E. Hedderman

Director, Performance Improvement

T. Hedges

Emergency Preparedness Manager

D. Holm

Plant Manager

F. Mis

General Supervisor, Radiation Protection

J. Pacher

Manager, Nuclear Engineering Services

J. Sullivan

Manager of Operations

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened 05000244/2009002-01

AV

Failure to Properly Lubricate Governor Linkage

(Section 4OA2)

Opened and Closed 05000244/2009002-02

FIN

Inadequate Risk Management Results in Loss

of Normal Control Room Annunciators

(Section 4OA3)

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

Document

UFSAR, Rev. 21

Procedure

O-22, Cold Weather Walkdown Procedure, Rev. 00500

Attachment

A-2

Section 1R04: Equipment Alignment

Documents

Component Cooling Water System Health Report, 1st Quarter, 2009

DBCOR 2004-0038, Miscellaneous Ginna Input Requested by Westinghouse Data Requests

Operating Experience Smart Sample, FY 2009-02, Negative Trend and Recurring Events

Involving Feedwater Systems, Rev. 0

Procedures

ATT-1.0, Attachment at Power CCW Alignment, Rev. 3

ATT-1.1, Attachment Normal CCW Flow, Rev. 0

S-30.5, Standby Auxiliary Feedwater Pump and Valve and Breaker, Rev. 34

S-30.9, Component Cooling Water Flow Path Verification, Rev. 2

Drawings

33013-1233, Condensate Low Pressure Feedwater Heaters, Rev.29

33013-1235, Condensate, Rev. 20

33013-1236, Feedwater, Sheet 1, Rev. 14

33013-1236, Feedwater, Sheet 2, Rev. 13

33013-1238, Standby Auxiliary Feedwater, Rev.25

33013-1239, Diesel Generator B, Rev. 21

33013-1245, Auxiliary Coolant Component Cooling Water, Rev. 31

33013-1246, Auxiliary Coolant Component Cooling Water, Sheet 1, Rev. 15

33013-1246, Auxiliary Coolant Component Cooling Water, Sheet 2, Rev. 12

33013-1252, Condensate, Rev. 23

Condition Reports

2006-7077

2006-7095

2006-7103

2006-7270

2007-5491

2008-0208

2008-0253

2008-3858

2008-4841

2008-4947

2009-1245

2009-1246

Work Orders

20501896

20600459

20602676

20701528

20702792

20703619

20703960

20706135

20800696

20800697

20800698

Section 1R05: Fire Protection

Document

Ginna Fire Protection Plan, Rev. 5

Procedures

FRP-6.0, Auxiliary Building Operating Floor, Rev. 6

FRP-29.0, Technical Support Center, Rev. 12

FRP-35.0, Standby Auxiliary Feedwater Building, Rev. 4

PT-13.4.29, Halon System Testing Relay Room/Computer Room, Rev. 02401

PT-13.4.35, Testing of Smoke Detection Zone Z-35 (Spent Fuel Area), Rev. 9

PT-13.11.4, Gamewell Smoke Detector Testing Zone Z25, Rev. 12

Attachment

A-3

PT-13.11.15, Testing of Fire Detection Zone Z-30 TSC Equipment Rooms-South, Rev. 10

PT-13.11.21, Gamewell Smoke Detector Testing Zone Z04, Rev. 1

PT-13.16.0, Star Corporation Heat Detector Zone Testing Zone Z05, Rev. 11

Section 1R06: Flood Protection Measures

Documents

I-DC-787-0428-13, Water Intrusion into RHR Pit from Auxiliary Building Suppression Systems,

Rev. 3

MPR-3084, Evaluation of Internal and External Flooding at R.E. Ginna Nuclear Power Plant,

Rev. 0

NUREG-0821, Integrated Plant Safety Assessment Systematic Evaluation Program, Rev. 0

PCR-2005-0037, Seismically Upgrade Reactor Water Makeup Tank and Monitor Tanks for RHR

Flooding Issues, Rev. 0

Drawing

33013-1271, Waste Disposal-Liquid RC Drain Tank P&ID, Rev. 13

Section 1R11: Licensed Operator Requalification

Documents

ANSI/ANS-3.4-1983, Medical Certification and Monitoring of Personnel Requiring Operator

Licenses for Nuclear Power Plants.

ANSI/ANS-3.5-1985, Nuclear Power Plant Simulators for Use in Operator Training

ES1213-05, Small Break Loss of Coolant Accident, Rev. 9

GSG-2.0, Simulator Testing, Rev. 2

OTG-12.0, Licensed Operator Requalification Training Schedule, Rev. 10

R.E. Ginna Operations PQW Qualification Matrix

R.E. Ginna 2009 Requalification Examination Sample Plan

R.E. Ginna Simulator Test Plan

TR-C.5.2, Licensed Operator Requalification Program, Rev. 35

Operating Experience:

OE-25273

OE-25091

OE-2008-0356

OE-2008-1212

Kewanunee 2007007/009

OE-2008-0144

OE-RIS2007-21

OE-2008-0024

Training Review Requests:

GNA-2008-281

GNA-2007-546

GNA-2007-559

GNA-LOR-2007-7

Training Change Orders:

GNA-LOR-2008-44

GNA-LOR-2007-157

GNA-LOR-2007-158

Attachment

A-4

Simulator Deficiency Reports:

SDR 2007-021

SDR 2007-036

SDR 2007-040

SDR 2007-081

SDR 2007-095

SDR 2007-131

SDR 2007-132

SDR 2008-066

SDR 2008-082

SDR 2008-086

SDR 2008-135

SDR 2008-153

Transient Tests:

14.4.8 BE-01, Manual Reactor Trip

14.4.8 BE-02, Trip of Feedwater Pumps

14.4.8 BE-03, Simultaneous Closure of Both MSIVs

14.4.8 BE-04, Simultaneous Trip of Both RCPs

14.4.8 BE-05, Single RCP Trip

14.4.8 BE-06, Main Turbine Trip

14.4.8 BE-07, Maximum Power Rate Ramp

14.4.8 BE-08, Maximum Size RCS Rupture W/Loss of All Offsite Power

14.4.8 BE-09, Maximum Unisolable Main Steam Line Rupture

14.4.8 BE-10, Slow RCS Depressurization Using PORV

Steady State and Computer Tests:

14.03.02, Computer Real Time Test

14.04.01, Operating Limits Monitoring

14.04.02, Normal Operations Acceptance Test

14.04.03.01, 100% Steady State Accuracy Test

14.04.03.02, 100% Power Steady State Drift Check

14.04.03.04, Initial Conditions Stability Check

14.04.04.01, NSSS - BOP Energy and Mass Balance

Procedures

CNG-TR-1.01-1000, Conduct of Training, Rev. 00200

CNG-SE-1.01-1001, Fitness for Duty Program, Rev. 00001

EPIP-2.18, Control Room Dose Assessment, Rev. 01600

OTG-2.2, Simulator Examination Instructions, Rev. 43

Condition Reports

2008-0393

2008-8713

2008-9753

2009-0146

2009-0232

2009-0203

2009-0297

Audits and Assessments:

Quarterly Report QPAR-2007-01-G

Quarterly Report QPAR-2007-02-G

Quarterly Report QPAR-2007-03-G

Quarterly Report QPAR-2007-04-G

Attachment

A-5

Quarterly Report QPAR-2008-01-G

Quarterly Report QPAR-2008-02-G

Quarterly Report QPAR-2008-03-G

Training and Qualifications Programs/TQS-08-01

Quality Performance Assessment Report 2007-0073

Quality Performance Assessment Report 2007-0083

Quality Performance Assessment Report 2008-0042

QPA Assessment Report 2007-0042

QPA Assessment Report 2007-0070

QPA Assessment Report 2007-0073

QPA Assessment Report 2007-0080

Section 1R12: Maintenance Effectiveness

Documents

Apparent Cause Evaluation for CR 2009-0129 (1/8/09)

Apparent Cause Evaluation for CR 2008-9624 (11/18/08)

CMIS Main Steam MR Train MSS01 Description and MR Functions

Control Building Ventilation, Ginna System Description, Chapter 22, Rev. 27

Control Building HVAC System (#71), System Health Report (Q1 - 2009)

Form MR5, Goal Determination for Control Room HVAC System CBV02, Rev. 2 (ID #: 2007-005)

Form MR5 Goal Determination for Main Steam MSS01, Rev. 1

Main Steam, Ginna System Description, Chapter 40, Rev. 12

Main Steam System (#81), System Health Report (Q1 - 2009)

MR Manager Scoping for CRV02A - CREATS Filtration Train A

MR Manager Scoping for CBV02 - Control Room Toxic Gas Monitors and Radiation Monitors

MR Manager Scoping for MSS01 - Main Steam Supply Header A

MR Status from Ginna Nuclear Engineering website (Revised 1/19/09)

Technical Basis for Continued Operability/Functionality CR-2008-7154, Attachment 5

TS 3.3.6 CREATS Actuation Instrumentation, Amendment 87 and Basis Document, Rev. 38

TS 3.7 Plant Systems, Amendment 80 and Revision Basis Document, Rev. 42

UFSAR Section 6.4.2 Control Room Ventilation System Design, Rev. 21

UFSAR Section 10.3 Main Steam System, Rev. 21

Procedures

CNG-AM-1.01-1023, Maintenance Rule Program, Rev. 00000

CNG-AM-1.01-2000, Scoping and Identification of Critical Components, Rev. 00200

Condition Reports

2009-1395

2008-9624

2008-8900

2008-7576

2008-7154

2008-5353

2008-4678

2007-3963

2009-1218

2009-0129

2008-8469

2008-1418

2007-8243

2007-2130

Work Orders

20806221

20806087

20805557

20804594

20803039

20803280

20803833

20900353

20900093

20404440

20706453

Attachment

A-6

Calculations

Ginna Calculation Note #67: Control Room Leak Rate as a Function of Control Room Leak Area

(R1213868; CALC-NOTE-67)

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

Documents

Integrated Work Schedule, Final Schedule, Week 344B

Procedures

CNG-OP-4.01-1000, Integrated Risk Management, Rev. 00100

M-94, Repair of RIS Alarm Panels in MCB, Rev. 008

O-6, Operations and Process Monitoring, Rev. 10200

O-6.13, Daily Surveillance Log, Rev. 16900

STP-O-12.2, Emergency Diesel Generator B, Rev. 00200

Condition Reports

2009-0253

2009-0278

2009-1647

2009-1651

Miscellaneous

Auto Log Entries for Equipment Log (OOS Only), 03/09/2009, 03/10/2009 and 03/12/2009

Auto Log Entries for Equipment Log Starting, 03/08/2009 to 03/12/2009 inclusive

Section 1R15: Operability Evaluations

Documents

DA-EE-92-084-21, Instrument Loop Performance Evaluation and Setpoint Verification ACC P936,

Rev. 2

Engineering Services Request 2009-0043, Past Operability of MOV 4007 and MOV 4008, Rev. 0,

February 13, 2009

IMC Part 9900: Technical Guidance for Operability Determinations and Functionality

Assessments

Proto Power Calculation 08-015, The Prevention of Vortices and Swirl at Intakes by Denny and

Young, Rev. A

Procedures

E-0, Reactor Trip or Safety Injection, Rev. 04200

E-3, Steam Generator Tube Rupture, Rev. 04500

O-6.13, Daily Surveillance Log, Rev. 16800

Drawing

33013-1237, Auxiliary Feedwater, Rev. 55

Condition Reports

2002-0525

2009-0242

2009-0437

2009-0738

2009-1305

2009-0903

Attachment

A-7

Section 1R18: Plant Modifications

Document

PCR 2008-0034, Installation of Rupture Disks Upstream of Service Water Thermal Relief Valves,

Rev. 0

Procedure

CNG-CM-1.01-1003, Design Engineering and Configuration Control, Rev. 00001

Drawing

33013-1250, Station Service Cooling Water Safety Related P&ID, Sheet 2, Rev. 36

Section 1R19: Post-Maintenance Testing

Procedures

GME-45-99-01, Electric Motor Inspection and Maintenance, Rev. 02101

STP-O-12.1, Emergency Diesel Generator A, Rev. 00401

STP-O-12.2, Emergency Diesel Generator B, Rev. 00301

STP-O-2.2QB, Residual Heat Removal Pump B Inservice Test, Rev. 00101

Condition Report

2009-1596

Work Orders

20805574

20807112

20800872

20900978

20805650

20805651

20805665

20900937

Section 1R22: Surveillance Testing

Documents

ACB 2000-0134, CCW Pump Test Flow

ACB 2000-0439, A CCW Pump Differential Pressure

Procedures

PT-36Q-C, Standby Auxiliary Feedwater Pump C - Quarterly, Rev. 05700

PT-16Q-T, Auxiliary Feedwater Turbine Pump - Quarterly, Rev. 05801

STP-O-2.8Q, Component Cooling Water Pump Quarterly Test, Rev. 00002

STP-O-12.2, Emergency Diesel Generator B, Rev. 00301

STP-O-22.2, Local Leak Rate Test of Personnel Hatch Door Seal, Rev. 00003

STP-O-16Q-B, Auxiliary Feedwater Pump B - Quarterly, Rev. 00300

Condition Reports

2009-0989

2008-9908

2008-9911

2006-7103

2009-1608

Attachment

A-8

Drawing

33013-1237, Auxiliary Feedwater P&ID, Rev. 55

Section 1EP6: Drill Evaluation

Documents

ES1213-05, Small Break Loss of Coolant Accident, Rev. 9

Section 4OA1: Performance Indicator Verification

Document

NEI 99-02, Nuclear Energy Institute Regulatory Assessment Performance Indicator Guideline,

Rev. 5, July 2007

Section 4OA2: Identification and Resolution of Problems

Documents

Category 1 Root Cause Analysis, CR-2008-9911, Turbine Driven Auxiliary Feedwater Pump

Failed to Develop Adequate Flow During Testing, dated January 9, 2009

EPRI Manual 1003084 Excerpts, Feedwater Pump Turbine Controls and Oil System

Maintenance Guide, dated December 2001

Ginna Probabilistic Risk Analysis Evaluation Request No. G1-2009-002, dated February 27, 2009

NUREG/CR-5857 Excerpts, Aging of Turbine Drives for Safety-Related Pumps in Nuclear Power

Plants, dated June 1995

Operating Experience Report - TDAFW Pump Failed to Develop Adequate Flow During Testing

Reptask P300158, Turbine Driven AFW Pump - Minor PM Inspection, M-11.5C

Standardized Plant Analysis Risk (SPAR) Model, Revision 3.45

Procedures

CNG-OP-1.01-1002, Conduct of Operability Determinations/Functionality Assessments,

Rev. 0000

M-11.5C, Auxiliary Feedwater Pump Minor Mechanical Inspection and Maintenance, Rev. 29,

dated February 27, 2006

PT-16Q-T, Auxiliary Feedwater Turbine Pump - Quarterly Rev. 05801

Condition Reports

2008-9911

2008-9956

Work Order 20602735

Section 4OA3: Followup of Events and Notices of Enforcement Discretion

Document

R.E. Ginna Emergency Action Level Technical Basis, Rev. 04400

Procedures

CNG-OP-4.01-1000, Integrated Risk Management, Rev. 00200

Attachment

A-9

M-94, Repair of RIS Alarm Panels in MCB, Rev. 8

Condition Reports

2009-0837

2009-0840

Work Order 20806014

Attachment

A-10

LIST OF ACRONYMS

ADAMS

Agencywide Documents Access and Management System

AFW

auxiliary feedwater

AV

apparent violation

CAP

corrective action program

CCDP

conditional core damage probability

CCW

component cooling water

CDF

core damage frequency

CR

condition report

EDG

emergency diesel generator

GINNA

R.E. Ginna Nuclear Power Plant

HX

heat exchanger

I&C

instrumentation and control

IMC

Inspection Manual Chapter

JPM

job performance measure

LOOP

loss of offsite power

MCB

main control board

MOV

motor-operated valve

NCV

non-cited violation

NEI

Nuclear Energy Institute

NRC

U.S. Nuclear Regulatory Commission

P&ID

piping and instrument drawings

PARS

Publicly Available Records

PCR

plant change record

PI

performance indicator

PM

preventive maintenance

PMT

post-maintenance testing

RBCCW

reactor building closed cooling water

RCP

reactor coolant pump

RG

regulatory guide

RHR

residual heat removal

SAFW

standby auxiliary feedwater

SDP

significance determination process

SPAR

standardized plant analysis risk

SRA

senior reactor analyst

SSC

system, structure, and component

SW

service water

TDAFW

turbine-driven auxiliary feedwater

TS

technical specification

UFSAR

updated final safety analysis report

UE

unusual event

WO

work order