ML093230675

IR 05000382-09-004; on July 8, 2009 Through October 7, 2009; Waterford Steam Electric Station, Unit 3; Operability Evaluations

ML093230675
Person / Time
Site:	Waterford
Issue date:	11/18/2009
From:	Clark J NRC/RGN-IV/DRP/RPB-E
To:	Kowalewski J Entergy Operations
References
IR-09-004
Download: ML093230675 (35)

See also: IR 05000382/2009004

Text

UNITED STATES

NU C LE AR RE G ULATO RY C O M M I S S I O N

R E GI ON I V

612 EAST LAMAR BLVD , SU ITE 400

AR LIN GTON , TEXAS 76011-4125

November 18, 2009

Joseph Kowalewski, Vice President, Operations

Entergy Operations, Inc.

Waterford Steam Electric Station, Unit 3

17265 River Road

Killona, LA 70057-3093

Subject: WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED

INSPECTION REPORT 05000382/2009-004

Dear Mr. Kowalewski:

On October 7, 2009, the U.S. Nuclear Regulatory Commission completed an inspection at your

Waterford Steam Electric Station, Unit 3. The enclosed integrated inspection report documents

the inspection findings, which were discussed on October 1, 2009, with you and other members

of your staff.

The inspections examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations and with the conditions of your license.

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

personnel.

This report documents one NRC identified finding of very low safety significance (Green). This

finding was determined to involve a violation of NRC requirements. However, because of the

very low safety significance and because it was entered into your corrective action program, the

NRC is treating this finding as a noncited violation, consistent with Section VI.A.1 of the NRC

Enforcement Policy. If you contest the violation or the significance of the noncited violation, you

should provide a response within 30 days of the date of this inspection report, with the basis for

your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,

Washington, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear

Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas,

76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,

Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Waterford Steam Electric

Station, Unit 3 facility. In addition, 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 IV, and the NRC

Resident Inspector at Waterford Steam Electric Station, Unit 3. The information you provide will

be considered in accordance with Inspection Manual chapter 0305.

Entergy Operations, Inc. -2-

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its

enclosure, will be available electronically for public inspection in the NRC Public Document

Room or from the Publicly Available Records component of NRCs document system (ADAMS).

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

Public Electronic Reading Room).

Sincerely,

/RA/

Jeffrey A. Clark, P.E.

Chief, Project Branch E

Division of Reactor Projects

Docket: 50-382

License: NPF-38

Enclosure:

NRC Inspection Report 05000382/2009004

w/Attachment: Supplemental Information

cc w/Enclosure:

Senior Vice President

Entergy Nuclear Operations

P.O. Box 31995

Jackson, MS 39286-1995

Senior Vice President and

Chief Operating Officer

Entergy Operations, Inc.

P.O. Box 31995

Jackson, MS 39286-1995

Vice President, Operations Support

Entergy Services, Inc.

P.O. Box 31995

Jackson, MS 39286-1995

Senior Manager, Nuclear Safety

and Licensing

Entergy Services, Inc.

P.O. Box 31995

Jackson, MS 39286-1995

Site Vice President

Waterford Steam Electric Station, Unit 3

Entergy Operations, Inc.

17265 River Road

Killona, LA 70057-0751

Entergy Operations, Inc. -3-

Director

Nuclear Safety Assurance

Entergy Operations, Inc.

17265 River Road

Killona, LA 70057-0751

General Manager, Plant Operations

Waterford 3 SES

Entergy Operations, Inc.

17265 River Road

Killona, LA 70057-0751

Manager, Licensing

Entergy Operations, Inc.

17265 River Road

Killona, LA 70057-3093

Chairman

Louisiana Public Service Commission

P.O. Box 91154

Baton Rouge, LA 70821-9154

Parish President Council

St. Charles Parish

P.O. Box 302

Hahnville, LA 70057

Director, Nuclear Safety & Licensing

Entergy, Operations, Inc.

440 Hamilton Avenue

White Plains, NY 10601

Louisiana Department of Environmental

Quality, Radiological Emergency Planning

and Response Division

P.O. Box 4312

Baton Rouge, LA 70821-4312

Chief, Technological Hazards Branch

FEMA Region VI

800 North Loop 288

Federal Regional Center

Denton, TX 76209

Entergy Operations, Inc. -4-

Electronic distribution by RIV:

Regional Administrator (Elmo.Collins@nrc.gov)

Deputy Regional Administrator (Chuck.Casto@nrc.gov)

DRP Director (Dwight.Chamberlain@nrc.gov)

DRP Deputy Director (Anton.Vegel@nrc.gov)

DRS Director (Roy.Caniano@nrc.gov)

DRS Deputy Director (Troy.Pruett@nrc.gov)

Senior Resident Inspector (Mark.Haire@nrc.gov)

Resident Inspector (Dean.Overland@nrc.gov)

Branch Chief, DRP/E (Jeff.Clark@nrc.gov)

Senior Project Engineer, DRP/E (Ray.Azua@nrc.gov)

WAT Site Secretary (Linda.Dufrene@nrc.gov)

Public Affairs Officer (Victor.Dricks@nrc.gov)

Branch Chief, DRS/TSB (Michael.Hay@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov)

Regional Counsel (Karla.Fuller@nrc.gov)

Congressional Affairs Officer (Jenny.Weil@nrc.gov)

OEMail Resource

Regional State Liaison Officer (Bill.Maier@nrc.gov)

NSIR/DPR/EP (Steve.LaVie@nrc.gov)

DRS STA (Dale.Powers@nrc.gov)

OEDO RIV Coordinator (Leigh.Trocine@nrc.gov)

ROPreports

File located: R:\_REACTORS\_WAT\2009\WAT 2009004 RP-DHO.doc

SUNSI Rev Compl. :Yes No ADAMS :Yes No Reviewer Initials

Publicly Avail  : Yes No Sensitive Yes : No Sens. Type Initials

RIV:SRI:DRP/E SPE/DRP/E C:DRS/EB1 C:DRS/EB2

DHOverland RAzua TRFarnholtz NFOKeefe

/RA/RAzua for /RA/ /RA/ /RA/

11/17/09 11/17/09 11/12/09 11/12/09

C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C:DRP/E

SGarchow MPShannon GEWerner JAClark

/RA/ /RA/ /RA/ /RA/RAzua for

11/17/2009 11/13/09 E11/11/09 11/17/09

OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000382

License: NFP-38

Report: 05000382/2009004

Licensee: Entergy Operations, Inc.

Facility: Waterford Steam Electric Station, Unit 3

Location: Hwy. 18

Killona, LA

Dates: July 8, 2009 through October 7, 2009

Inspectors: D. Overland, Senior Resident Inspector

R. Egli, Branch Chief, TTC

R. Hickok, Senior Reactor Technology Instructor, TTC

G. Replogle, Senior Project Engineer, RIV

M. Chambers, Resident Inspector, Cooper Nuclear Station

P. Jayroe, Project Engineer, RIV

T. Buchanan, Project Engineer, RIV

S. Anderson, General Engineer, HQ

L. Carson II, Senior Health Physicist

Approved By: Jeff Clark, Chief, Project Branch E

Division of Reactor Projects

-1- Enclosure

SUMMARY OF FINDINGS

IR 05000382/2009004; July 8, 2009 through October 7, 2009; Waterford Steam Electric Station,

Unit 3; Operability Evaluations.

The report covered a 3-month period of inspection by resident inspectors and announced

baseline inspections by regional based inspectors. One Green noncited violation of significance

was identified. The significance of most findings is indicated by their color (Green, White,

Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination Process.

Findings for which the significance determination process does not apply may be Green or be

assigned a severity level after NRC management review. The NRC's program for overseeing

the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor

Oversight Process, Revision 4, dated December 2006.

A. NRC-Identified Findings and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green. The inspectors identified a Green non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation. The technical specifications require all four

channels (A, B, C, and D) of local power density, departure from nucleate boiling ratio,

and reactor coolant flow instruments to be operable when in Mode 1. These Channel B

instruments require an input from the Channel B log power instrument, which was

previously declared inoperable. With the Channel B log power instrument inoperable,

the Channel B local power density, departure from nucleate boiling ratio, and reactor

coolant flow instruments should also have been declared inoperable. The licensee

entered this finding in their corrective action program as condition reports CR-WF3-

2009-4401 and CR-WF3-2009-4407.

The failure to either trip or bypass the inoperable channels within one hour was more

than minor because it affected the configuration control attribute of the mitigating

systems cornerstone. Specifically, deliberate operator action was required to ensure that

proper reactor protection system coincidence and reliability were maintained. Also, if left

uncorrected, the potential existed for Channel B reactor protective trips to be

inadvertently removed while at power. The failure to meet the technical specifications

was considered to be of very low safety significance (Green), since there was no actual

loss of safety function. This finding has a cross-cutting aspect in the decision-making

component of the human performance area because the licensee failed to verify the

validity of underlying assumptions and identify unintended consequences of failing to

comply with technical specification 3.3.1 by declaring the log power Channel B

inoperable and not placing local power density, departure from nucleate boiling ratio,

and reactor coolant flow instrument channels in either bypass or trip condition (H.1.b).

(Section 1R15)

B. Licensee-Identified Violations

None

-2- Enclosure

REPORT DETAILS

Summary of Plant Status

The plant began the inspection period on July 8, 2009, at 100 percent power and remained at

approximately 100 percent power for the rest of the inspection period.

1. REACTOR SAFETY

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

Emergency Preparedness

1R01 Adverse Weather Protection (71111.01)

.1 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

Since thunderstorms with potential tornados and high winds were forecast in the vicinity

of the facility for October 4, 2009, the inspectors reviewed the licensees overall

preparations/protection for the expected weather conditions. The inspectors evaluated

the licensee staffs documented preparations against the sites procedures and

determined that the staffs actions were adequate. During the inspection, the inspectors

focused on plant-specific design features and the licensees procedures used to respond

to specified adverse weather conditions. The inspector's evaluated operator staffing and

accessibility of controls and indications for those systems required to control the plant.

Additionally, the inspectors reviewed the Updated Final Safety Analysis Report and

verified that operator actions were appropriate as specified by plant-specific procedures.

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

These activities constitute completion of one readiness for impending adverse weather

condition sample as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignments (71111.04)

.1 Partial Walkdown

a. Inspection Scope

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

systems:

• July 22, 2009: Chemical volume control Train A

• August 12, 2009: Emergency feedwater Train A

• August 13, 2009: Low pressure safety injection Train B

• August 18, 2009: Emergency feedwater Train AB

• September 15, 2009: High pressure safety injection system Train A

-3- Enclosure

The inspectors selected these systems based on their risk significance relative to the

reactor safety cornerstones at the time they were inspected. The inspectors attempted

to identify any discrepancies that could affect the function of the system, and, therefore,

potentially increase risk. The inspectors reviewed applicable operating procedures,

system diagrams, Updated Final Safety Analysis Report, technical specification

requirements, administrative technical specifications, outstanding work orders, condition

reports, and the impact of ongoing work activities on redundant trains of equipment in

order to identify conditions that could have rendered the systems incapable of

performing their intended functions. The inspectors also walked down accessible

portions of the systems to verify system components and support equipment were

aligned correctly and operable. The inspectors examined the material condition of the

components and observed operating parameters of equipment to verify that there were

no obvious deficiencies. The inspectors also verified that the licensee had properly

identified and resolved equipment alignment problems that could cause initiating events

or impact the capability of mitigating systems or barriers and entered them into the

corrective action program with the appropriate significance characterization. Specific

documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of five partial system walkdown samples as

defined in Inspection Procedure 71111.04-05.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05)

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns that were focused on availability,

accessibility, and the condition of firefighting equipment in the following risk-significant

plant areas:

• July 21, 2009: Reactor auxiliary building fire Zones 8B, 8C, 11, and 12

• July 22, 2009: Reactor auxiliary building fire Zones 33, 35, 38, and 39

• July 30, 2009: Reactor auxiliary building fire Zones 3, 5, and 6

• August 3, 2009: Fire Zones Roof E and Roof W

• August 11, 2009: Reactor auxiliary building fire Zone 16

• August 18, 2009: Reactor auxiliary building fire Zones 33, 35, 36, 37, 38, and 39

• August 19, 2009: Reactor auxiliary building fire Zone 32

• August 20, 2009: Reactor auxiliary building fire Zones 2, Roof E, and Roof W

• August 23, 2009: Reactor auxiliary building fire Zones 11, 12,13, 8B, and 8C

• August 24, 2009: Reactor auxiliary building fire Zones 15, 16, 17, 18, 19, 20,

and 21

The inspectors reviewed areas to assess if licensee personnel had implemented a fire

protection program that adequately controlled combustibles and ignition sources within

-4- Enclosure

the plant; effectively maintained fire detection and suppression capability; maintained

passive fire protection features in good material condition; and had implemented

adequate compensatory measures for out of service, degraded or inoperable fire

protection equipment, systems, or features, in accordance with the licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk

as documented in the plants Individual Plant Examination of External Events with later

additional insights, their potential to affect equipment that could initiate or mitigate a plant

transient, or their impact on the plants ability to respond to a security event. Using the

documents listed in the attachment, the inspectors verified that fire hoses and

extinguishers were in their designated locations and available for immediate use; that

fire detectors and sprinklers were unobstructed, that transient material loading was

within the analyzed limits; and fire doors, dampers, and penetration seals appeared to

be in satisfactory condition. The inspectors also verified that minor issues identified

during the inspection were entered into the licensees corrective action program.

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

These activities constitute completion of ten quarterly fire-protection inspection samples

as defined in Inspection Procedure 71111.05-05.

b. Findings

No findings of significance were identified.

.2 Annual Fire Protection Drill Observation (71111.05A)

a. Inspection Scope

On September 23, 2009, the inspectors observed a fire brigade activation as the result

of a simulated fire at feed heater drain Pump C. The observation evaluated the

readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee

staff identified deficiencies, openly discussed them in a self-critical manner at the drill

debrief, and took appropriate corrective actions. Specific attributes evaluated were:

(1) proper wearing of turnout gear and self-contained breathing apparatus; (2) proper

use and layout of fire hoses; (3) employment of appropriate fire fighting techniques;

(4) sufficient firefighting equipment brought to the scene; (5) effectiveness of fire brigade

leader communications, command, and control; (6) search for victims and propagation of

the fire into other plant areas; (7) smoke removal operations; (8) utilization of pre

planned strategies; (9) adherence to the preplanned drill scenario; and (10) drill

objectives.

These activities constitute completion of one annual fire-protection inspection sample as

defined in Inspection Procedure 71111.05-05.

b. Findings

No findings of significance were identified.

-5- Enclosure

1R11 Licensed Operator Requalification Program (71111.11)

a. Inspection Scope

On August 4, 2009, the inspectors observed a crew of licensed operators in the plants

simulator to verify that operator performance was adequate, evaluators were identifying

and documenting crew performance problems, and training was being conducted in

accordance with licensee procedures. The inspectors evaluated the following areas:

• Licensed operator performance

• Crews clarity and formality of communications

• Crews ability to take timely actions in the conservative direction

• Crews prioritization, interpretation, and verification of annunciator alarms

• Crews correct use and implementation of abnormal and emergency procedures

• Control board manipulations

• Oversight and direction from supervisors

• Crews ability to identify and implement appropriate technical specification actions

and emergency plan actions and notifications

The inspectors compared the crews performance in these areas to pre-established

operator action expectations and successful critical task completion requirements.

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

These activities constitute completion of one quarterly licensed-operator requalification

program sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q)

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk

significant systems:

• August 11, 2009: Seal leakage on chemical volume control charging pumps

• September 3, 2009: Review of operating experience smart sample FY 2009-01,

Inspection of electrical connections for motor control center, circuit breakers and

interfaces

-6- Enclosure

The inspectors reviewed events such as where ineffective equipment maintenance has

resulted in valid or invalid automatic actuations of engineered safeguards systems and

independently verified the licensee's actions to address system performance or condition

problems in terms of the following:

• Implementing appropriate work practices

• Identifying and addressing common cause failures

• Scoping of systems in accordance with 10 CFR 50.65(b)

• Characterizing system reliability issues for performance

• Charging unavailability for performance

• Trending key parameters for condition monitoring

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

• Verifying appropriate performance criteria for structures, systems, and

components classified as having an adequate demonstration of performance

through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as

requiring the establishment of appropriate and adequate goals and corrective

actions for systems classified as not having adequate performance, as described

in 10 CFR 50.65(a)(1)

The inspectors assessed performance issues with respect to the reliability, availability,

and condition monitoring of the system. In addition, the inspectors verified maintenance

effectiveness issues were entered into the corrective action program with the appropriate

significance characterization. Specific documents reviewed during this inspection are

listed in the attachment.

These activities constitute completion of two quarterly maintenance effectiveness

samples as defined in Inspection Procedure 71111.12-05.

b. Findings

No findings of significance were identified.

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

a. Inspection Scope

The inspectors reviewed licensee personnel's evaluation and management of plant risk

for the maintenance and emergent work activities affecting risk-significant and safety-

related equipment listed below to verify that the appropriate risk assessments were

performed prior to removing equipment for work:

• July 29, 2009: Scheduled elective maintenance outage for containment fan

coolers Train B to calibrate containment fan cooler Header B CCW return

temperature control valve solenoid Valve CC-835B

-7- Enclosure

• August 3, 2009: Scheduled surveillance of reactor protection system Channel A

• September 9, 2009: Scheduled activity to remove high pressure safety injection

Pump AB from high pressure safety injection Train A alignment and align high

pressure safety injection Pump A to Train A

• September 11, 2009: Emergent maintenance to replace station Battery AB,

Cell 31 with a spare cell due to degraded cell voltage

The inspectors selected these activities based on potential risk significance relative to

the reactor safety cornerstones. As applicable for each activity, the inspectors verified

that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)

and that the assessments were accurate and complete. When licensee personnel

performed emergent work, the inspectors verified that the licensee personnel promptly

assessed and managed plant risk. The inspectors reviewed the scope of maintenance

work, discussed the results of the assessment with the licensee's probabilistic risk

analyst or shift technical advisor, and verified plant conditions were consistent with the

risk assessment. The inspectors also reviewed the technical specification requirements

and inspected portions of redundant safety systems, when applicable, to verify risk

analysis assumptions were valid and applicable requirements were met. Specific

documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four maintenance risk assessments and

emergent work control inspection samples as defined in Inspection

Procedure 71111.13-05.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope

The inspectors reviewed the following issues:

• July 14, 2009: Low individual cell voltage on vital 125 vdc station Battery AB

Cell 39

• August 11, 2009: Unplanned load variations during emergency diesel generator

Train A surveillance

• August 12, 2009: Emergency diesel generator Train A Relay EG EREL 2342(J)

found out of calibration during surveillance

• August 20, 2009: Channel B local power density, departure from nucleate boiling

ratio, and reactor coolant flow instruments, when Channel B log power

instrument was inoperable

-8- Enclosure

The inspectors selected these potential operability issues based on the risk-significance

of the associated components and systems. The inspectors evaluated the technical

adequacy of the evaluations to ensure that technical specification operability was

properly justified and the subject component or system remained available such that no

unrecognized increase in risk occurred. The inspectors compared the operability and

design criteria in the appropriate sections of the technical specifications and Updated

Safety Analysis Report to the licensees evaluations, to determine whether the

components or systems were operable. Where compensatory measures were required

to maintain operability, the inspectors determined whether the measures in place would

function as intended and were properly controlled. The inspectors determined, where

appropriate, compliance with bounding limitations associated with the evaluations.

Additionally, the inspectors also reviewed a sampling of corrective action documents to

verify that the licensee was identifying and correcting any deficiencies associated with

operability evaluations. Specific documents reviewed during this inspection are listed in

the attachment.

These activities constitute completion of four operability evaluations inspection samples

as defined in Inspection Procedure 71111.15-05

b. Findings

Introduction: The inspectors identified a Green non-cited violation of technical

specification 3.3.1, Reactor Protective Instrumentation. The technical specifications

require all four channels (A, B, C, and D) of local power density, departure from nucleate

boiling ratio, and reactor coolant flow instruments to be operable when in Mode 1.

These Channel B instruments require an input from the Channel B log power instrument,

which was previously declared inoperable. With the Channel B log power instrument

inoperable, the Channel B local power density, departure from nucleate boiling ratio, and

reactor coolant flow instruments should also have been declared inoperable.

Description: On Aug 20, 2009, the inspector observed the performance of procedure

MI-003-126, Revision 14, Core Protection Calculator Functional. During the

performance of the test procedure, the inspector noted that CPC Channel B high log

power trip was bypassed. The inspector asked why technical specification 3.3.1 had not

been entered due to the inoperable log power Channel B instrument. Technical

specification 3.3.1, Reactor Protective Instrumentation, requires that the reactor

protective instrumentation channels and bypasses contained in Table 3.3-1 be operable

in accordance with the requirements of the table. Table 3.3-1 requires all four channels

of local power density (LPD), departure from nucleate boiling ratio (DNBR), and reactor

coolant flow instruments to be operable in Mode 1.

Log power Channel B provides a high log power automatic bypass removal signal for

LPD, DNBR, and reactor coolant flow instrumentation channels. Technical specification 3.3.1, Table 3.3-1 requires the high log power bypass shall be automatically removed

when reactor power is greater than or equal to 10-4% of rated thermal power. When in

Mode 1, reactor power is greater than 10-4% of rated thermal power. The inspectors

determined that when a log power instrument is out of service, the automatic removal of

the high log power bypass function is inoperable and thus the associated protective

channels of LPD, DNBR, and reactor coolant flow are also inoperable.

-9- Enclosure

The log power Channel B instrument was originally declared inoperable on Sept 1, 2008.

The operability determination concluded that since the plant was in Mode 4, only two log

power channels were required, therefore entry into technical specification 3.3.1 was not

required. On Sept 9, 2008, the plant entered Mode 2 with log power Channel B still

inoperable. The operability was not revised to reflect the change in plant conditions. In

accordance with technical specification 3.3.1, operators should have taken action to

place the associated LPD, DNBR, and reactor coolant flow protective channels to either

bypass or trip within one hour.

On Aug 22, 2009, after considering the inspectors question, the licensee declared LPD

Channel B and DNBR Channel B inoperable, and placed both instruments in bypass.

During a subsequent control room tour, the inspector verified that LPD and DNBR were

bypassed, however noticed that reactor coolant flow Channel B had not been bypassed.

The inspector asked the shift manager if technical specification 3.3.1, Table 3.3-1

notation (C) affected any other trips. Upon further assessment, operations personnel

determined that reactor coolant low flow was also affected and declared steam

generator flow Channel B to be inoperable, as well.

Analysis: The failure to either trip or bypass the inoperable channels within one hour

was more than minor because it affected the configuration control attribute of the

mitigating systems cornerstone. Specifically, deliberate operator action was required to

ensure that proper reactor protection system coincidence and reliability were maintained.

Also, if left uncorrected, the potential existed for Channel B reactor protective trips to be

inadvertently removed while at power. The failure to meet the technical specifications

was considered to be of very low safety significance (Green), since there was no actual

loss of safety function. This finding has a cross-cutting aspect in the decision-making

component of the human performance area because the licensee failed to verify the

validity of underlying assumptions and identify unintended consequences of failing to

comply with technical specification 3.3.1 by declaring the log power Channel B

inoperable and not placing DNBR, LPD, and reactor coolant flow channels in either

bypass or trip condition (H.1.b).

Enforcement: Technical specification 3.3.1, Reactor Protective Instrumentation,

requires all four channels of LPD, DNBR, and reactor coolant flow to be operable and

able to have the high log power bypass automatically removed when reactor power is

greater than or equal to 10-4% percent of rated thermal power. Contrary to this, on

September 9, 2008, the licensee did not comply with the limiting condition for operation

action statement for technical specification 3.3.1 which states, the inoperable channel is

placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The plant remained in

this condition until August 22, 2009. This violation has been determined to be of very

low safety significance and was entered into their corrective action program in condition

reports CR-WF3-2009-4401 and CR-WF3-2009-4407. Therefore, this violation is being

treated as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC

Enforcement Policy.

- 10 - Enclosure

1R18 Plant Modifications (71111.18)

a. Inspection Scope

The inspectors reviewed the following temporary/permanent modifications to verify that

the safety functions of important safety systems were not degraded:

• August 26, 2009: Permanent modification of containment vacuum relief valves

such that once the valves are automatically opened, they remain open until

manually closed.

• August 7, 2009: Temporary modification to revise the setpoint for the reactor

coolant Pump 2A upper thrust bearing high temperature alarm to reduce

nuisance alarms in the control room.

• September 14, 2009: Temporary modification to replace station Battery AB, Cell

31 with a new cell. The old Cell 31 was left in place and jumpered around, while

the new Cell 31 was installed at the end of the battery rack.

The inspectors reviewed the temporary modification and the associated safety

evaluation screening against the system design bases documentation, including the

Updated Final Safety Analysis Report and the technical specifications, and verified that

the modification did not adversely affect the system operability/availability. The

inspectors also verified that the installation and restoration were consistent with the

modification documents and that configuration control was adequate. Additionally, the

inspectors verified that the temporary modification was identified on control room

drawings, appropriate tags were placed on the affected equipment, and licensee

personnel evaluated the combined effects on mitigating systems and the integrity of

radiological barriers.

The inspectors reviewed key affected parameters associated with energy needs,

materials/replacement components, timing, heat removal, control signals, equipment

protection from hazards, operations, flow paths, pressure boundary, ventilation

boundary, structural, process medium properties, licensing basis, and failure modes for

the modification listed below. The inspectors verified that modification preparation,

staging, and implementation did not impair emergency/abnormal operating procedure

actions, key safety functions, or operator response to loss of key safety functions;

postmodification testing will maintain the plant in a safe configuration during testing by

verifying that unintended system interactions will not occur, systems, structures and

components performance characteristics still meet the design basis, the

appropriateness of modification design assumptions, and the modification test

acceptance criteria will be met; and licensee personnel identified and implemented

appropriate corrective actions associated with permanent plant modifications. Specific

documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three samples for temporary and permanent

plant modifications as defined in Inspection Procedure 71111.18-05

- 11 - Enclosure

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing (71111.19)

a. Inspection Scope

The inspectors reviewed the following postmaintenance activities to verify that

procedures and test activities were adequate to ensure system operability and functional

capability:

• June 23, 2009: Replacement of high pressure safety injection Pump B Tyco time

delay relay following the failure of the relay to start the pump during a routine

surveillance test

• July 23, 2009: Replacement of seal package on chemical volume control

charging Pump B to reduce reactor coolant system unidentified leakage

• July 29, 2009: Scheduled elective maintenance calibration of containment fan

cooler Header B CCW return temperature control valve solenoid Valve CC-835B

• August 4, 2009: Corrective maintenance to repair the actuator for steam

generator SG1 main steam atmospheric dump Valve MS-116A

• August 11, 2009: Scheduled preventative maintenance to clean, inspect, and

test emergency diesel generator Train A Relay EG EREL 2342(J)

• September 9, 2009: Scheduled preventative maintenance to replace the

pulsation dampener and perform motor maintenance on chemical volume control

charging Pump AB

• September 14, 2009: Emergent maintenance to replace station Battery AB,

Cell 31 with a spare cell, due to degraded voltage on the cell

• September 29, 2009: Scheduled preventative maintenance to check the

overcurrent trip on the breaker for non-nuclear safety return header isolation

Valve CC-562.

The inspectors selected these activities based upon the structure, system, or

component's ability to affect risk. The inspectors evaluated these activities for the

following (as applicable):

• The effect of testing on the plant had been adequately addressed; testing was

adequate for the maintenance performed

• Acceptance criteria were clear and demonstrated operational readiness; test

instrumentation was appropriate

- 12 - Enclosure

The inspectors evaluated the activities against the technical specifications, the Updated

Final Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and

various NRC generic communications to ensure that the test results adequately ensured

that the equipment met the licensing basis and design requirements. In addition, the

inspectors reviewed corrective action documents associated with postmaintenance tests

to determine whether the licensee was identifying problems and entering them in the

corrective action program and that the problems were being corrected commensurate

with their importance to safety. Specific documents reviewed during this inspection are

listed in the attachment.

These activities constitute completion of eight postmaintenance testing inspection

sample(s) as defined in Inspection Procedure 71111.19-05.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report, procedure

requirements, and technical specifications to ensure that the six surveillance activities

listed below demonstrated that the systems, structures, and/or components tested were

capable of performing their intended safety functions. The inspectors either witnessed or

reviewed test data to verify that the significant surveillance test attributes were adequate

to address the following:

• Preconditioning

• Evaluation of testing impact on the plant

• Acceptance criteria

• Test equipment

• Procedures

• Jumper/lifted lead controls

• Test data

• Testing frequency and method demonstrated technical specification operability

• Test equipment removal

• Restoration of plant systems

• Fulfillment of ASME Code requirements

- 13 - Enclosure

• Updating of performance indicator data

• Engineering evaluations, root causes, and bases for returning tested systems,

structures, and components not meeting the test acceptance criteria were correct

• Reference setting data

• Annunciators and alarms setpoints

The inspectors also verified that licensee personnel identified and implemented any

needed corrective actions associated with the surveillance testing.

• August 6, 2009: Safety related electrical Bus 3A undervoltage relay calibration

• August 10, 2009: Emergency diesel generator Train A surveillance

• August 20, 2009: Core protection calculator Train B surveillance

• August 22, 2009: Plant protection system Channel B surveillance

• August 24, 2009: Emergency diesel generator and subgroup relays Train B

• September 14, 2009: High pressure safety injection Train AB

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

These activities constitute completion of six surveillance testing inspection samples as

defined in Inspection Procedure 71111.22-05.

b. Findings

No findings of significance were identified.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation (71114.06)

.1 Training Observations

a. Inspection Scope

The inspectors observed a training evolution for licensed operators on September 17,

2009, which required emergency plan implementation by a licensee operations crew.

This evolution was planned to be evaluated and included in performance indicator data

regarding drill and exercise performance. The inspectors observed event classification

and notification activities performed by the crew. The inspectors also attended the

postevolution critique for the scenario. The focus of the inspectors activities was to note

any weaknesses and deficiencies in the crews performance and ensure that the

licensee evaluators noted the same issues and entered them into the corrective action

program. As part of the inspection, the inspectors reviewed the scenario package and

other documents listed in the attachment.

These activities constitute completion of one sample as defined in Inspection

Procedure 71114.06-05.

- 14 - Enclosure

b. Findings

No findings of significance were identified.

2. RADIATION SAFETY

Cornerstone: Occupational and Public Radiation Safety

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual and

collective radiation exposures ALARA. The inspector used the requirements in 10 CFR

Part 20 and the licensees procedures required by technical specifications as criteria for

determining compliance. The inspector interviewed licensee personnel and reviewed:

• Current 3-year rolling average collective exposure

• Five outage work activities scheduled during the inspection period and

associated work activity exposure estimates which were likely to result in the

highest personnel collective exposures

• Site-specific trends in collective exposures, plant historical data, and source-term

measurements

• Five work activities of highest exposure significance completed during the last

outage

• ALARA work activity evaluations, exposure estimates, and exposure mitigation

requirements

• Intended versus actual work activity doses and the reasons for any

inconsistencies

• Person-hour estimates provided by maintenance planning and other groups to

the radiation protection group with the actual work activity time requirements

• Post-job (work activity) reviews

• Assumptions and basis for the current annual collective exposure estimate, the

methodology for estimating work activity exposures, the intended dose outcome,

and the accuracy of dose rate and man-hour estimates

• Method for adjusting exposure estimates, or re-planning work, when unexpected

changes in scope or emergent work were encountered

• Exposure tracking system

- 15 - Enclosure

• Exposures of individuals from selected work groups

• Records detailing the historical trends and current status of tracked plant source

terms and contingency plans for expected changes in the source term due to

changes in plant fuel performance issues or changes in plant primary chemistry

• Declared pregnant workers during the current assessment period, monitoring

controls, and the exposure results

• Self-assessments, audits, and special reports related to the ALARA program

since the last inspection

• Resolution through the corrective action process of problems identified through

post-job reviews and post-outage ALARA report critiques

The inspector completed 11 of the required 15 samples and 5 of the optional samples as

defined in IP 71121.02-05.

4. OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151)

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the data submitted by the licensee for the second

quarter of 2009 performance indicators for any obvious inconsistencies prior to its public

release in accordance with Inspection Manual Chapter 0608, Performance Indicator

Program.

This review was performed as part of the inspectors normal plant status activities and,

as such, did not constitute a separate inspection sample.

b. Findings

No findings of significance were identified.

.2 Safety System Functional Failures

a. Inspection Scope

The inspectors sampled licensee submittals for the Safety System Functional Failures

performance indicator for the period from the second quarter of 2008 through the second

quarter of 2009. To determine the accuracy of the performance indicator data reported

during those periods, performance indicator definitions and guidance contained in NEI

Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5,

and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73" definitions

and guidance were used. The inspectors reviewed the licensees operator narrative

logs, operability assessments, maintenance rule records, maintenance work orders,

- 16 - Enclosure

issue reports, event reports and NRC Integrated Inspection reports for the period

beginning the second quarter of 2008 through the second quarter of 2009 to validate the

accuracy of the submittals. The inspectors also reviewed the licensees issue report

database to determine if any problems had been identified with the performance

indicator data collected or transmitted for this indicator and none were identified.

Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one safety system functional failures sample as

defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.3 Mitigating Systems Performance Index - Emergency ac Power System

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance

Index (MSPI) - Emergency ac Power System performance for the period from the

second quarter of 2008 through the second quarter of 2009. To determine the accuracy

of the performance indicator data reported during those periods, performance indicator

definitions and guidance contained in NEI Document 99-02, Regulatory Assessment

Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the

licensees operator narrative logs, mitigating systems performance index derivation

reports, issue reports, event reports and NRC integrated inspection reports for the period

beginning the second quarter of 2008 through the second quarter of 2009 to validate the

accuracy of the submittals. The inspectors reviewed the mitigating systems performance

index component risk coefficient to determine if it had changed by more than 25 percent

in value since the previous inspection, and if so, that the change was in accordance with

applicable NEI guidance. The inspectors also reviewed the licensees issue report

database to determine if any problems had been identified with the performance

indicator data collected or transmitted for this indicator and none were identified.

Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one mitigating systems performance index

emergency ac power system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.4 Mitigating Systems Performance Index - Cooling Water Systems

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance

Index - Cooling Water Systems performance for the period from the second quarter of

2008 through the second quarter of 2009. To determine the accuracy of the

performance indicator data reported during those periods, performance indicator

definitions and guidance contained in NEI Document 99-02, Regulatory Assessment

- 17 - Enclosure

Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the

licensees operator narrative logs, issue reports, mitigating systems performance index

derivation reports, event reports and NRC integrated inspection reports for the period

beginning the second quarter of 2008 through the second quarter of 2009 to validate the

accuracy of the submittals. The inspectors reviewed the mitigating systems performance

index component risk coefficient to determine if it had changed by more than 25 percent

in value since the previous inspection, and if so, that the change was in accordance with

applicable NEI guidance. The inspectors also reviewed the licensees issue report

database to determine if any problems had been identified with the performance

indicator data collected or transmitted for this indicator and none were identified.

Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one mitigating systems performance index

cooling water system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.16 Occupational Exposure Control Effectiveness (OR01)

a. Inspection Scope

The inspector sampled licensee submittals for the Occupational Radiological

Occurrences performance indicator for the first quarter of 2009 through the third quarter

of 2009. To determine the accuracy of the performance indicator data reported during

those periods, performance indicator definitions and guidance contained in NEI

Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5,

were used. The inspector reviewed the licensees assessment of the performance

indicator for occupational radiation safety to determine if indicator related data was

adequately assessed and reported. To assess the adequacy of the licensees

performance indicator data collection and analyses, the inspector discussed with

radiation protection staff, the scope and breadth of its data review, and the results of

those reviews. The inspector independently reviewed electronic dosimetry dose rate

and accumulated dose alarm and dose reports and the dose assignments for any

intakes that occurred during the time period reviewed to determine if there were

potentially unrecognized occurrences. The inspector also conducted walkdowns of

numerous locked high and very high radiation area entrances to determine the adequacy

of the controls in place for these areas.

These activities constitute completion of the occupational radiological occurrences

sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

- 18 - Enclosure

.17 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual

Radiological Effluent Occurrences (PR01)

a. Inspection Scope

The inspector sampled licensee submittals for the Radiological Effluent Technical

Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences

performance indicator for the first quarter of 2009 through the third quarter of 2009. To

determine the accuracy of the performance indicator data reported during those periods,

performance indicator definitions and guidance contained in NEI Document 99-02,

Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The

inspector reviewed the licensees issue report database and selected individual reports

generated since this indicator was last reviewed to identify any potential occurrences

such as unmonitored, uncontrolled, or improperly calculated effluent releases that may

have impacted offsite dose. The inspector reviewed gaseous effluent summary data and

the results of associated offsite dose calculations for selected dates during the third

quarter of 2009 to determine if indicator results were accurately reported. The inspector

also reviewed the licensees methods for quantifying gaseous and liquid effluents and

determining effluent dose. Additionally, the inspector reviewed the licensees historical

10 CFR Part 50.75(g) file and selectively reviewed the licensees analysis for discharge

pathways resulting from a spill, leak, or unexpected liquid discharge focusing on those

incidents which occurred over the last few years.

These activities constitute completion of the radiological effluent technical

specifications/offsite dose calculation manual radiological effluent occurrences sample

as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152)

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

Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical

Protection

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of

this report, the inspectors routinely reviewed issues during baseline inspection activities

and plant status reviews to verify that they were being entered into the licensees

corrective action program at an appropriate threshold, that adequate attention was being

given to timely corrective actions, and that adverse trends were identified and

addressed. The inspectors reviewed attributes that included: the complete and

accurate identification of the problem; the timely correction, commensurate with the

safety significance; the evaluation and disposition of performance issues, generic

implications, common causes, contributing factors, root causes, extent of condition

- 19 - Enclosure

reviews, and previous occurrences reviews; and the classification, prioritization, focus,

and timeliness of corrective. Minor issues entered into the licensees corrective action

program because of the inspectors observations are included in the attached list of

documents reviewed.

These routine reviews for the identification and resolution of problems did not constitute

any additional inspection samples. Instead, by procedure, they were considered an

integral part of the inspections performed during the quarter and documented in

Section 1 of this report.

b. Findings

No findings of significance were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific

human performance issues for follow-up, the inspectors performed a daily screening of

items entered into the licensees corrective action program. The inspectors

accomplished this through review of the stations daily corrective action documents.

The inspectors performed these daily reviews as part of their daily plant status

monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings of significance were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensees corrective action program and

associated documents to identify trends that could indicate the existence of a more

significant safety issue. The inspectors focused their review on repetitive equipment

issues, but also considered the results of daily corrective action item screening

discussed in Section 4OA2.2, above, licensee trending efforts, and licensee human

performance results. The inspectors nominally considered the 6-month period of

January 2009 through July 2009, for a review of Operating Experience Smart Sample:

OpESS FY2009-02, A Negative trend and Recurring Events Involving feedwater

systems as it applies to the emergency feedwater system.

The inspectors also included issues documented outside the normal corrective action

program in major equipment problem lists, repetitive and/or rework maintenance lists,

departmental problem/challenges lists, system health reports, quality assurance

audit/surveillance reports, self-assessment reports, and Maintenance Rule assessments.

The inspectors compared and contrasted their results with the results contained in the

licensees corrective action program trending reports. Corrective actions associated with

- 20 - Enclosure

a sample of the issues identified in the licensees trending reports were reviewed for

adequacy.

These activities constitute completion of one single semi-annual trend inspection sample

as defined in Inspection Procedure 71152-05.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

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

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

Steam Electric Station 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.

4OA6 Meetings

Exit Meeting Summary

On September 18, 2009, the team presented the inspection results to Mr. J. Kowalewski,

Vice President, Operations, and other members of his staff who acknowledged the findings.

The team confirmed that proprietary information was not provided or examined during the

inspection.

On October 1, 2009, the inspectors presented the inspection results to Mr. Joe Kowalewski, and

other members of the licensee staff. The licensee acknowledged the issues presented. The

inspector asked the licensee whether any materials examined during the inspection should be

considered proprietary. No proprietary information was identified.

- 21 - Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

M. Adams, Supervisor, System Engineering

S. Anders, Manager, Plant Security

C. Arnone, Plant Manager

J. Brawley, ALARA Supervisor, Radiation Protection

B. Briner, Technical Specialist IV, Componet Engineering

K. Christian, Director, Nuclear Safety Assurance

K. Cook, Manager, Operations

L. Dauzat, Supervisor, Radiation Protection

D. Dufrene; Technician, Radiation Protection

C. Fugate, Assistant Manager, Operations

M. Haydel, Supervisor, Programs and Components

J. Kowalewski, Vice President of Operations

J. Lewis, Manager, Emergency Preparedness

B. Lindsey, Manager, Maintenance

M. Mason, Senior Licensing Specialist, Licensing

W. McKinney, Manager, Corrective Action and Assessments

C. Miller, Lead Supervisor, Radiation Protection

R. Murillo, Manager, Licensing

K. Nicholas, Director, Engineering

B. Piluti, Manager, Radiation Protection

J. Polluck, Engineer, Licensing

R. Putnam, Manager, Programs and Components

S. Ramzy; Specialist, Radiation Protection

J. Ridge, Manager, Quality Assurance

J. Solaski, Quality Assurance Auditor

J. Williams, Senior Licensing Specialist, Licensing

NRC Personnel

S. Anderson, General Engineer, HQ

T. Buchanan, Project Engineer, RIV

L. Carson II, Senior Health Physicist

M. Chambers, Resident Inspector, Cooper Nuclear Station

R. Egli, Branch Chief, Technical Training Center

R. Hickok, Senior Reactor Technology Instructor, Technical Training Center

P. Jayroe, Project Engineer, RIV

G. Replogle, Senior Project Engineer, RIV

A-1 Attachment

LIST OF ITEMS OPENED AND CLOSED

Opened and Closed

Failure to Follow Technical Specification Requirements for

05000382/2009004-1 NCV

Reactor Protective Instrumentation

LIST OF DOCUMENTS REVIEWED

Section 1RO1: Adverse Weather Protection

CONDITION REPORTS

CR-WF3-1998-00710

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

OP-901-521 Sever Weather and Flooding 301

Section 1RO4: Equipment Alignment

CONDITION REPORTS

CR-WF3-2009-0607 CR-WF3-2009-0737 CR-WF3-2009-1189 CR-WF3-2009-1624

CR-WF3-2009-2869

WORK ORDERS

190714

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

OP-903-045 Emergency Feedwater Flow Path Lineup Verification 5

OP-009-008 Safety Injection System 26

OP-002-005 Chemical and Volume Control 28

SD-CVC Chemical and Volume Control System Description 6

SD-SI Safety Injection System Description 6

A-2 Attachment

Section 1RO5: Fire Protection

CONDITION REPORTS

CR-WF3-2009-04034 CR-WF3-2009-04035 CR-WF3-2009-04060

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

UNT-005-013 Fire Protection Program 10

OP-009-004 Fire Protection 305

MM-004-424 Building Fire Hose Station Inspection and Hose 10

Replacement

MM-007-010 Fire Extinguisher Inspection and Extinguisher Replacement 302

FP-001-014 Duties of a Fire Watch 14

FP-001-015 Fire Protection Impairments 302

DBD-018 Appendix R/Fire Protection

FP-001-015 Fire Protection Impairments 302

FP-001-018 Pre-fire Plan Strategies, Development, And Revision 300

UNT-007-006 Housekeeping 301

EN-DC-161 Control of Combustibles 003

UNT-007-060 Control of Loose Items 302

UNT-005-013 Fire Protection Program 010

Engineering Calculations F91-044 01

Engineering Calculations F91-019 0

Section 1R11: Licensed Operator Requalification Program

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

Simulator Scenario Number E-70

Simulator Scenario Number E-125

OP-901-201 Steam Generator Level Control System Malfunction 009

A-3 Attachment

OP-902-000 Standard Post Trip Actions 010

OP-902-008 Safety function Recovery Procedure 015

OP-901-110 Pressurizer Level Control Malfunction 005

OP-901-311 Loss of Train B Safety Bus 302

OP-901-102 CEA or CEDMCS Malfunciton 300

OP-902-001 Reactor Trip Recovery 011

OP-902-002 Loss of Coolant Accident Recovery Procedure 012

Section 1R12: Maintenance Effectiveness

CONDITION REPORTS

CR-WF3-2007-3497 CR-WF3-2008-4306 CR-WF3-2008-3836 CR-WF3-2009-0506

CR-WF3-2008-4189 CR-WF3-2008-4611 CR-WF3-2009-1190 CR-WF3-2009-4131

CR-WF3-2008-4297 CR-WF3-2008-4765 CR-WF3-2009-2862 CR-WF3-2009-3810

CR-WF3-2008-1072 CR-WF3-2008-2410 CR-WF3-2008-2352 CR-WF3-2008-4332

CR-WF3-2008-1796 CR-WF3-2008-2810 CR-WF3-2008-2579 CR-WF3-2008-5045

CR-WF3-2008-1807 CR-WF3-2008-3363 CR-WF3-2008-4127 CR-WF3-2008-5273

CR-WF3-2008-2066 CR-WF3-2008-2346 CR-WF3-2008-4173 CR-WF3-2009-0955

CR-WF3-2009-1200 CR-WF3-2009-1284 CR-WF3-2009-4015 CR-WF3-2009-4324

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

EN-DC-206 Maintenance Rule 1

NUMARC 93-01 Industry Guideline for Monitoring the Effectiveness of 3

maintenance at Nuclear Power Plants

Section 1R13: Maintenance Risk Assessment and Emergent Work Controls

WORK ORDERS

51802942 52039753 0019397401 52192184

197692

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

OI-037-000 Operations Risk Management Guideline 300

EN-WM-101 On-Line Work Management Process 1

A-4 Attachment

W2.502 Configuration risk Management Program 000

OP-100-010 Equipment Out of Service 303

OP-903-107 Plant Protection System channel A & B & C & D 303

Functional Test

OP-903-030 Safety Injection Pump Operability Verification 18

OP-009-008 Safety Injection System 26

OP-006-003 125 VDC Electrical Distribution 301

ME-003-200 Station Battery Bank and Charger (Weekly) 301

ME-003-210 Station Battery Bank and Charger (Quarterly) 12

Section 1R15: Operability Evaluations

CONDITION REPORTS

CR-WF3-2009-4466 CR-WF3-2009-4163 CR-WF3-2009-4395 CR-WF3-2009-4401

CR-WF3-2009-4407 CR-WF3-2009-3540 CR-WF3-2009-4139 CR-WF3-2009-3448

CR-WF3-2009-3557

WORK ORDERS

5180191 52038533

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

EN-OP-104 Operability Determinations 4

ME-003-200 Station Battery Bank and Charger (Weekly) 301

ME-003-210 Station Battery Bank and Charger (Quarterly) 12

OP-006-003 125 Vdc Electrical Distribution 301

OP-006-001 Plant Distribution System 305

MI-003-126 Core Protection Calculator Functional 14

SD-PPS Plant Protection System Description 0

OP-903-107 Plant Protection System Channel A, B, C, D, Functional Test 303

TSTF-324 Correct logarithmic power vs. RTP 1

A-5 Attachment

Section 1R18: Plant Modifications

CONDITION REPORTS

CR-WF3-2009-3399

WORK ORDERS

203111 197692

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION /

DATE

EN-DC-136 Temporary Modifications 4

EC NO: 706 Modification of containment relief valves 0

EN-WM-105 Implement EC 706 2/3/2007

EN-WM-105 Implement EC 15451 2/3/2007

ME-004-213 Battery Intercell Connections 14

16496 Temporary Modification

Section 1R19: Postmaintenance Testing

CONDITION REPORTS

CR-WF3-2009-3102 CR-WF3-2009-4304 CR-WF3-2009-3448 CR-WF3-2009-4139

CR-WF3-2009-4766

WORK ORDERS

199029 51802942 52039753 0019397401

199977 188048 52040097 52038057

51523543 201698 52194563 197692

5180191

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

OP-903-030 Safety Injection Pump Operability Verification 15

OP-903-068 Emergency Diesel Generator Operability and Subgroup 303

Relay Operability Verification

OP-009-008 Safety Injection System 25

A-6 Attachment

Section 1R19: Postmaintenance Testing

CONDITION REPORTS

CR-WF3-2009-3102 CR-WF3-2009-4304 CR-WF3-2009-3448 CR-WF3-2009-4139

CR-WF3-2009-4766

WORK ORDERS

199029 51802942 52039753 0019397401

199977 188048 52040097 52038057

51523543 201698 52194563 197692

5180191

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

OP-903-118 Primary Auxiliaries Quarterly IST Valve Tests 18

OP-903-037 Containment Cooling Fan Operability Verification 5

OP-903-119 Secondary Auxiliaries Quarterly IST Valve Tests 9

OP-903-120 Containment and Miscellaneous Systems Quarterly IST 9

Valve Tests

OP-903-003 Charging Pump Operability Check 301

ME-004-213 Battery Intercell Connections 14

OP-903-118 Primary Auxiliaries Quarterly IST Valve Tests 18

ME-007-002 Molded Case Circuit Breakers 15

SD-CC Component Cooling Water and Auxiliary Component 7

Cooling Water System Description

STA-001-005 Leakage testing of Air and Nitrogen Accumulators for Safety 304

Related Valves

Section 1R22: Surveillance Testing

CONDITION REPORTS

CR-WF3-2009-04053 CR-WF3-2009-04072 CR-WF3-2009-04073 CR-WR3-2009-4395

CR-WF3-2009-4401 CR-WF3-2009-4203 CR-WF3-2008-4163 CR-WR3-2009-4466

A-7 Attachment

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

ME-003-318 G.E. Undervoltage Relay Model 121AV55C 303

OP-009-002 Emergency Diesel Generator Start Evaluation [Data Sheet] 310

OP-009-002 Diesel Generator Start Running Log 310

OP-903-068 Emergency Diesel Generator A Surveillance Test

OP-903-068 Emergency Diesel Generator and Subgroup Relay 303

Operability Verification - Train B

OP-903-030 Safety Injection Pump Operability Verification 18

OP-009-008 Safety Injection System 26

OP-903-107 Plant Protection System Channel B Functional Test 303

MI-003-126 Core Protection Calculator Functional 014

Section 1EP6: Drill Evaluation

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

EP-001-001 Recognition and Classification of Emergency Conditions 22

EP-001-030 Site Area Emergency 300

EP-001-040 General Emergency 300

Scenario DEP 2007-02

Section 2OS2: ALARA Planning and Controls

PROCEDURES/DOCUMENTS

NUMBER TITLE REVISION

EN-RP-102 Radiological Control 0

EN-RP-105 Radiation Work Permits 4

EN-RP-106 Radiological Survey Documentation 2

EN-RP-110 ALARA Program 2

EN-RP-141 Job Coverage 6

EN-DIR-RP-002 Radiation Protection Performance Indicator 0

A-8 Attachment

AUDITS, SELF-ASSESSMENTS, AND SURVEILLANCES

NUMBER TITLE DATE

QA-14/15-2009-WF3-1 Radiation Protection/Radwaste Audit

Quality Oversight Observations May 2008

RADIATION WORK PERMITS

RWP# RWP DESCRIPTION

2008-0511 1R15 S/G Primary Side Eddy Current Testing Inspection and Repair

2008-0610 1R Scaffolding

2008-0631 1R15 Alloy 600 Mitigation Activities Pressurizer/Hot Legs (Weld Overlay)

2008-0702 Reactor Disassembly

2008-0705 Reactor Reassembly

CONDITION REPORTS

CR-WF3-2008-1699 CR-WF3-2008-1776 CR-WF3-2008-1793 CR-WF3-2008-1946

CR-WF3-2008-1989 CR-WF3-2008-2027 CR-WF3-2008-2347 CR-WF3-2008-4495

CR-WF3-2009-4959 CR-WF3-2009-4969

MISCELLANEOUS

TITLE DATE

Waterford 3 Refuel Reactor Coolant System Dose Equivalent Iodine September 10, 2009

Reactor Coolant System Cleanup Flow Chart

5-Year ALARA Plan

Refueling Outage 15 Report

Failed Fuel Shutdown Mitigation Plan

Section 4OA1: Performance Indicator Verification

PROCEDURES

NUMBER TITLE REVISION

NEI 99-02 Regulatory Assessment Performance Indicator Guideline 5

EN-LI-114 Performance Indicator Process 4

A-9 Attachment

Section 4OA2: Identification and Resolution of Problems

CONDITION REPORTS

CR-WF3-2008-4000 CR-WF3-2008-4748 CR-WF3-2008-5793 CR-WF3-2009-0089

CR-WF3-2009-0570 CR-WF3-2009-1416 CR-WF3-2009-2604 CR-WF3-2009-3294

CR-WF3-2009-0754 CR-WF3-2009-1446 CR-WF3-2009-2706 CR-WF3-2009-3651

CR-WF3-2009-0770 CR-WF3-2009-2136 CR-WF3-2009-

WORK ORDERS

178225 51665138

PROCEDURES

NUMBER TITLE REVISION

EFW System Health Report 1st Quarter 2009 4/30/09

A-10 Attachment