ML070450251

IR 05000382-06-005, on 10/08/2006-12/31/2006, Waterford Steam Electric Station, Unit 3, Postmaintenance Testing; Refueling and Other Outage Activities, Emergency Plan Biennial Program Inspection

ML070450251
Person / Time
Site:	Waterford
Issue date:	02/13/2007
From:	Clark J NRC/RGN-IV/DRP/RPB-E
To:	Walsh K Entergy Operations
References
IR-06-005
Download: ML070450251 (43)

See also: IR 05000382/2006005

Text

February 13, 2007

Kevin Walsh

Vice President Operations

Waterford 3

Entergy Operations, Inc.

17265 River Road

Killona, LA 70066-0751

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

INSPECTION REPORT 05000382/2006005

Dear Mr. Walsh:

On December 31, 2006, the NRC completed an inspection at your Waterford Steam Electric

Station, Unit 3. The enclosed report documents the inspection findings, which were discussed

on January 16, 2007, with you and other members of your staff.

This inspection examined activities conducted under your license as they relate to safety and

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

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

personnel.

This report documents two self-revealing and one NRC identified findings of very low safety

significance (Green). These findings were determined to involve violations of NRC

requirements. Additionally, a licensee-identified violation, which was determined to be of very

low safety significance is listed in this report. However, because of the very low safety

significance and because they were entered into your corrective action program, the NRC is

treating these violations as noncited violations (NCVs), consistent with Section VI.A.1 of the

NRC Enforcement Policy. If you contest any NCV in this report, you should provide a response

within 30 days of the date of this inspection report, with the basis for your denial, to the

U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC

20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission,

Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of

Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the

NRC Resident Inspector at the Waterford Steam Electric Station, Unit 3, facility.

Entergy Operations, Inc. -2-

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

enclosures, and your response, if any, will be made available electronically for public inspection

in the NRC Public Document Room or from the Publicly Available Records (PARS) component

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

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

Sincerely,

/RA/

Jeff A. Clark, Chief

Project Branch E

Division of Reactor Projects

Docket: 50-382

License: NPF-38

Enclosure: NRC Inspection Report 050000382/2006005

w/Attachment: Supplemental Information

cc w/enclosure:

Executive Vice President and

Chief Operating Officer

Entergy Operations, Inc.

P.O. Box 31995

Jackson, MS 39286-1995

Vice President, Operations Support

Entergy Operations, Inc.

P.O. Box 31995

Jackson, MS 39286-1995

General Manager, Plant Operations

Waterford 3 SES

Entergy Operations, Inc.

17265 River Road

Killona, LA 70057-3093

Manager, Licensing

Entergy Operations, Inc.

17265 River Road

Killona, LA 70057-3093

Chairman

Louisiana Public Service Commission

P.O. Box 91154

Baton Rouge, LA 70825-1697

Entergy Operations, Inc. -3-

Director, Nuclear Safety Assurance

Entergy Operations, Inc.

17265 River Road

Killona, LA 70057-3093

Richard Penrod, Senior Environmental

Scientist, State Liaison Officer

Office of Environmental Services

Northwestern State University

Russsell Hall, Room 201

Natchitoches, LA 71497

Parish President

Council

St. Charles Parish

P.O. Box 302

Hahnville, LA 70057

Chairperson

Denton Field Office

Chemical and Nuclear Preparedness and Protection Division

Office of Infrastructure Protection

Preparedness Directorate

Dept. of Homeland Security

800 North Loop 288

Federal Regional Center

Denton, TX 76201-3698

Entergy Operations, Inc. -4-

Electronic distribution by RIV:

Regional Administrator (BSM1)

DRP Director (ATH)

DRS Director (DDC)

DRS Deputy Director (RJC1)

Senior Resident Inspector (GFL1)

Branch Chief, DRP/E (ZKD)

Senior Project Engineer, DRP/E (VGG)

Team Leader, DRP/TSS (RLN1)

RITS Coordinator (MSH3)

Regional State Liaison Officer (WAM)

NSIR/DPR/EPD (JTJ1)

NSIR/DPR/EPD (REK)

DRS STA (DAP)

D. Cullison, OEDO RIV Coordinator (DGC)

ROPreports

WAT Site Secretary (AHY)

SUNSI Review Completed: _JAC____ ADAMS: / Yes G No Initials: __JAC____

/ Publicly Available G Non-Publicly Available G Sensitive / Non-Sensitive

R:\_REACTORS\_WAT\2006\WT2006-05RP-GFL.wpd

RI:DRP/E SRI:DRP/E C:DRS/EB1 C:DRS/OB C:DRS/EB2

DHOverland GFLarkin WBJones ATGody LJSmith

T-JAC T-JAC /RA/ /RA/ /RA/

2/5/07 2/5/07 2/5/07 2/8/07 2/8/07

C:PSB C:DRP/E

MPShannon JAClark

/RA/ /RA/

2/8/07 2/13/07

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

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 50-382

License No.: NPF-38

Report No.: 05000382/2006005

Licensee: Entergy Operations, Inc.

Facility: Waterford Steam Electric Station, Unit 3

Location: Hwy. 18

Killona, Louisiana

Dates: October 8 through December 31, 2006

Inspectors: G. Larkin, Senior Resident Inspector

D. Overland, Resident Inspector

J. Kirkland, Project Engineer, Project Branch E

T. McKernon, Senior Operations Engineer, Operations Branch

L. Carson II, Senior Health Physics Inspector, Plant Support

Branch

G. George, Reactor Inspector, Engineering Branch 1

P. Elkmann, Emergency Preparedness Inspector, Operations

Branch

Approved By: Jeff A. Clark, Chief, Project Branch E

ATTACHMENTS: Supplemental Information

Enclosure

SUMMARY OF FINDINGS

IR05000382/2006-005; 10/08/2006-12/31/2006; Waterford Steam Electric Station, Unit 3;

Postmaintenance Testing; Refueling and Other Outage Activities; Emergency Plan Biennial

Program Inspection

The report covered a 3-month period of inspection by resident inspectors, a project engineer, a

senior operations engineer, a senior health physics inspector, a reactor inspector, and an

emergency preparedness inspector. The inspectors identified two Green findings, which were

noncited violations, and one apparent violation with potential safety significance greater than

Green. 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 NRCs program for overseeing the safe

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

Oversight Process, Revision 3, dated July 2000.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Initiating Events

• Green. A self-revealing violation of very low safety significance of Technical

Specification 6.8.1.a was identified for an inadequate procedure for installing a

bolted joint that provided structural support for the pressurizer. Specifically, the

installation procedure required applying 8750 ft-lbs torque to make up a bolted

joint. Following corrective actions, the licensee discovered that the break away

torque on several bolts exceeded 13,400 ft-lbs. The improper bolt tensioning

resulted in failure of 1 of 16 bolts and the partial cracking of 3 other bolts that

potentially could affect the pressurizers function in a safe shutdown earthquake

event. The licensee has since replaced all pressurizer skirt bolting and installed

the bolting to an approved torque specification.

This finding is more than minor because if left uncorrected it could have become

a more safety significant concern. The finding was associated with the

equipment performance attribute of the Initiating Events cornerstone, and it

affected the cornerstone objective to limit the likelihood of those events that

upset plant stability and challenge critical safety functions during power

operations. This finding was determined to have very low safety significance

because a seismic event would not have resulted in a loss-of-coolant accident

that exceeded the Technical Specification limit for reactor coolant system

leakage. Therefore, this issue screened out in Phase 1 of the Manual

Chapter 0609 Significance Determination Process, because there was no

actual loss of safety function (Section 1R20).

Enclosure

Cornerstone: Mitigating Systems

• Green. A self-revealing violation of very low safety significance of 10 CFR

Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the

failure to implement effective corrective actions to prevent recurrence of a

significant condition adverse to quality. Specifically, on multiple occasions

Valve SI-405B failed to stroke open while attempting to place shutdown cooling

Train B in service. This violation of Appendix B, Criterion XVI, is being treated as

a noncited violation and was entered into the licensees corrective action

program.

This finding is greater than minor because it affects the Mitigating Systems

cornerstone attribute of equipment operability, availability, and reliability of

systems that respond to initiating events. This finding was evaluated using the

significance determination process and was determined to be a finding of very

low safety significance because, in each condition identified, it did not represent

an actual loss of a safety function. The inspectors also determined that the

cause of the condition had crosscutting aspects associated with the corrective

action program component in the problem identification and resolution area.

This assessment was based on the fact that the licensee failed to thoroughly

evaluate the problem such that the resolutions addressed the causes and

therefore, corrective actions were inadequate to prevent repetition

(Section 1R19).

Cornerstone: Emergency Preparedness

• Green. The inspector identified a noncited violation of 10 CFR 50.54(q) for

failure to conduct during 2005 an offsite drill involving a simulated contaminated

individual with provision for participation by local medical support services as

required by the licensees emergency plan. The licensees failure to conduct the

drill is a performance deficiency because the licensee identified the drills

postponement in October 2005, but did not appropriately reschedule the drill. In

addition, the licensee did not request NRC approval to deviate from this

emergency plan requirement.

This finding is greater than minor because a degraded proficiency in providing

appropriate medical treatment for a contaminated individual has a potential

impact on the safety of licensee employees and the public. The finding is of very

low safety significance because the licensee failed to conduct only one required

drill during the inspection period January 2005 through December 2006, and the

drill was not appropriately rescheduled with NRC approval. This finding is a

noncited violation of 10 CFR 50.54(q) and 10 CFR Part 50, Appendix E, IV, F.1.

The licensee has entered this issue into their corrective action system as

Condition Report 2006-4429 (Section 1EP5).

-2- Enclosure

B. Licensee-Identified Violations

Violations of very low safety significance, which were identified by the licensee have

been reviewed by the inspectors. Corrective actions taken or planned by the licensee

have been entered into the licensees corrective action program. These violations and

corrective actions are listed in Section 4OA7 of this report.

-3- Enclosure

REPORT DETAILS

Summary of Plant Status: The plant was operated at approximately 100 percent power from

October 8 to November 22, 2006, when reactor power was reduced to 85 percent power to

comply with a Technical Specification requirement to reduce power to compensate for a main

steam safety valve lift point setting found outside of its allowed tolerance. Reactor power was

increased to 100 percent power on November 23, 2006. On November 25, 2006, operators

commenced a plant shutdown for Refueling Outage 14. Operators restarted the reactor plant

on December 26, 2006 and reached 100 percent power on December 30, 2006, and remained

at 100 percent power through the end of the report period.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection (71111.01)

.1 Readiness For Seasonal Susceptibilities

a. Inspection Scope

The inspectors completed a review of the licensees readiness of seasonal

susceptibilities involving low seasonal temperatures and high winds. The inspectors:

(1) reviewed plant procedures, the Updated Final Safety Analysis Report, and Technical

Specifications to ensure that operator actions defined in adverse weather procedures

maintained the readiness of essential systems; (2) walked down portions of the three

systems listed below to ensure that adverse weather protection features (heat tracing,

space heaters and weatherized enclosures) were sufficient to support operability,

including the ability to perform safe shutdown functions; (3) evaluated operator staffing

levels to ensure the licensee could maintain the readiness of essential systems required

by plant procedures; and (4) reviewed the corrective action program to determine if the

licensee identified and corrected problems related to adverse weather conditions.

• December 13, 2006: Main Steam System, Firewater System, and Emergency

Feedwater System

Documents reviewed by the inspectors included Operations Procedure OP-901-521,

Severe Weather and Flooding, Revision 4-3, Operations Procedure OP-002-007,

Freeze Protection and Temperature Maintenance, Revision 11, and Design Basis

Document W3-DBD-003, Emergency Feedwater System.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

-4- Enclosure

1R04 Equipment Alignment (71111.04)

.1 Partial Walkdown

a. Inspection Scope

The inspectors: (1) walked down portions of the two below listed risk important systems

and reviewed plant procedures and documents to verify that critical portions of the

selected systems were correctly aligned; and (2) compared deficiencies identified during

the walk down to the licensees Updated Final Safety Analysis Report and corrective

action program to ensure problems were being identified and corrected.

• October 10, 2006: Chemical and Volume Control System Train A

• December 11, 2006: Shutdown Cooling System Train A

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed two samples.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05)

.1 Quarterly Inspection

a. Inspection Scope

The inspectors walked down the six below listed plant areas to assess the material

condition of active and passive fire protection features and their operational lineup and

readiness. The inspectors: (1) verified that transient combustibles and hot work

activities were controlled in accordance with plant procedures; (2) observed the

condition of fire detection devices to verify they remained functional; (3) observed fire

suppression systems to verify they remained functional and that access to manual

actuators was unobstructed; (4) verified that fire extinguishers and hose stations were

provided at their designated locations and that they were in a satisfactory condition;

(5) verified that passive fire protection features (electrical raceway barriers, fire doors,

fire dampers, steel fire proofing, penetration seals, and oil collection systems) were in a

satisfactory material condition; (6) verified that adequate compensatory measures were

established for degraded or inoperable fire protection features and that the

compensatory measures were commensurate with the significance of the deficiency;

and (7) reviewed the Updated Final Safety Analysis Report to determine if the licensee

identified and corrected fire protection problems.

• October 10, 2006: Fire Zones RAB 16, 23, 36, 39, and Fuel Handling Building

• October 16, 2006: Fire Zones RAB 2, 16, 18 and 19

-5- Enclosure

• October 18, 2006: Fire Zones RAB 2, 8A, 36, 37, and ultimate heat sink Train A

• November 21, 2006: Fire Zones RAB 1E, 6, 7, 11, 12, 13

• November 30, 2006: Fire Zones RAB 2, 8B, 23, 31, 39

• December 5, 2006: Fire Zones RAB 15, 19, 20, 21, 22, Containment Building,

Cooling Tower A

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed six samples.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures (71111.06)

.2 Semi-annual Internal Flooding

a. Inspection Scope

The inspectors: (1) reviewed the Updated Final Safety Analysis Report, the flooding

analysis, and plant procedures to assess seasonal susceptibilities involving external

flooding; (2) reviewed the Updated Final Safety Analysis Report and corrective action

program to determine if the licensee identified and corrected flooding problems;

(3) inspected underground bunkers/manholes to verify the adequacy of (a) sump

pumps, (b) level alarm circuits, (c) cable splices subject to submergence, and

(d) drainage for bunkers/manholes; (4) verified that operator actions for coping with

flooding can reasonably achieve the desired outcomes; and (5) walked down the one

below listed areas to verify the adequacy of: (a) equipment seals located below the

floodline, (b) floor and wall penetration seals, (c) watertight door seals, (d) common

drain lines and sumps, (e) sump pumps, level alarms, and control circuits, and

(f) temporary or removable flood barriers.

• December 5, 2006: Reactor Containment Building

The inspectors reviewed calculation MN(Q)-6-4, Water Level Inside Containment,

dated November 2, 1978.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

-6- Enclosure

1R11 Licensed Operator Requalification Program (71111.11)

.1 Biennial Inspection

a. Inspection Scope

The inspector reviewed the annual operating examination test results for 2006. Since

this was the first half of the biennial requalification cycle, the licensee was not required

to administer a written examination. These results were assessed to determine if they

were consistent with NUREG 1021, Operator Licensing Examination Standards for

Power Reactors, guidance and Manual Chapter 0609, Appendix I, Operator

Requalification Human Performance Significance Determination Process,

requirements. This review included examination of test results for a total of 42 licensed

operators, a total of 9 crews, which included: shift-standing senior operators, staff

senior operators, shift-standing reactor operators, and staff reactor operators. All crews

and individuals passed the requalification examinations.

The inspector completed one sample.

b. Findings

No findings of significance were identified.

.2 Quarterly Inspection

a. Inspection Scope

On November 6-7, 2006, the inspectors observed training of senior reactor operators

and reactor operators to identify deficiencies and discrepancies in the training, to assess

operator performance, and to assess the evaluators critique. The training scenario

involved a simulated plant shutdown exercise in preparation for the plant shutdown on

November 26, 2006 for a plant refueling outage.

Documents reviewed by the inspectors included:

• Operations Procedure OP-010-005, Plant Shutdown, Revision 5

• Emergency Planning Procedure EP-001-001, Recognition and Classification of

Emergency Conditions, Revision 21

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

-7- Enclosure

1R12 Maintenance Effectiveness (71111.12)

a. Inspection Scope

The inspectors reviewed the two below listed Maintenance Rule scoped systems that

have displayed performance problems to: (1) verify the appropriate handling of

structure, system, and component performance or condition problems; (2) verify the

appropriate handling of degraded structure, system, and component functional

performance; (3) evaluate the role of work practices and common cause problems; and

(4) evaluate the handling of structure, system, and component issues reviewed under

the requirements of the Maintenance Rule, 10 CFR Part 50 Appendix B, and the

Technical Specifications.

• Essential Chill Water System

• Startup Transformers (Offsite Power)

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed two samples.

b. Findings

No findings of significance were identified.

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

.1 Risk Assessment and Management of Risk

a. Inspection Scope

The inspectors reviewed the four below listed assessment activities to verify:

(1) performance of risk assessments when required by 10 CFR 50.65 (a)(4) and

licensee procedures prior to changes in plant configuration for maintenance activities

and plant operations; (2) the accuracy, adequacy, and completeness of the information

considered in the risk assessment; (3) that the licensee recognizes, and/or enters as

applicable, the appropriate licensee-established risk category according to the risk

assessment results and licensee procedures; and (4) the licensee identified and

corrected problems related to maintenance risk assessments.

• October 31, 2006: Planned maintenance activities on start up transformer

Train B

• November 29, 2006: Planned maintenance and operational activities during

reactor coolant system midloop conditions

• December 7, 2006: Planned maintenance activities to restore emergency diesel

generator Train B following the integrated diesel test

• December 18, 2006: Planned maintenance and operational activities during

reactor coolant system midloop conditions

-8- Enclosure

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed four samples.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope

The inspectors: (1) reviewed plant status documents such as operator shift logs,

emergent work documentation, deferred modifications, and standing orders to

determine if an operability evaluation was warranted for degraded components;

(2) referred to the Updated Final Safety Analysis Report and design-basis documents to

review the technical adequacy of licensee operability evaluations; (3) evaluated

compensatory measures associated with operability evaluations; (4) determined

degraded component impact on any Technical Specifications; (5) used the Significance

Determination Process to evaluate the risk significance of degraded or inoperable

equipment; and (6) verified that the licensee has identified and implemented appropriate

corrective actions associated with degraded components.

• October 19, 2006: Operability evaluation addressing pressurizer heater capacity

following a surveillance test as described in Condition Report 2006-3125

• November 11, 2006: Operability evaluation addressing containment fan cooler

Train B flow control valve failing to control at intermediate flow rates as described

in Condition Report 2006-3357

• December 22, 2006: Operability evaluation addressing steam Generator 32 tube

sheet plugging as described in Condition Report 2006-4510

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed three samples.

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing (71111.19)

a. Inspection Scope

The inspectors selected the six below listed postmaintenance test activities of risk

significant systems or components. For each item, the inspectors: (1) reviewed the

applicable licensing basis and/or design-basis documents to determine the safety

functions; (2) evaluated the safety functions that may have been affected by the

maintenance activity; and (3) reviewed the test procedure to ensure it adequately tested

the safety function that may have been affected. The inspectors either witnessed or

-9- Enclosure

reviewed test data to verify that acceptance criteria were met, plant impacts were

evaluated, test equipment was calibrated, procedures were followed, jumpers were

properly controlled, the test data results were complete and accurate, the test

equipment was removed, the system was properly realigned, and deficiencies during

testing were documented. The inspectors also reviewed the Updated Final Safety

Analysis Report to determine if the licensee identified and corrected problems related to

postmaintenance testing.

• October 11, 2006: Planned maintenance on chemical volume control charging

Pump B

• October 12, 2006: Planned maintenance for reactor trip circuit breaker

Number 7

• November 1, 2006: Planned maintenance for start-up transformer Train B

6.9 kV circuit breaker

• November 26, 2006: Emergent maintenance on reactor coolant Loop 1

shutdown cooling suction inside containment isolation Valve SI-405B

• November 29, 2006: Planned maintenance on excore start-up Channel 2

• December 12, 2006: Planned maintenance on reactor coolant Loop 1 shutdown

cooling suction outside containment isolation Valve SI-407B

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed six samples.

b. Findings

Introduction. A self-revealing violation of very low safety significance of 10 CFR Part 50,

Appendix B, Criterion XVI, Corrective Action, was identified for the failure to implement

effective corrective actions to prevent recurrence of a significant condition adverse to

quality. Specifically, on multiple occasions Valve SI-405B failed to stroke open while

attempting to place shutdown cooling Train B in service. The root cause analysis

determined that the design of SI-405A(B) fluid reservoir was intolerant to operation with

low pressure/level. This condition caused inadequate hydraulic pump discharge

pressure that resulted in the failure of Valve SI-405B to stroke open. This violation of

Appendix B, Criterion XVI, is being treated as a noncited violation and was entered into

the licensees corrective action program.

Description. On November 26, 2006, while in Mode 4, station operators performed

Operating Procedure OP-903-033, Cold Shutdown IST Valve Test to stroke open

Valve SI-405B in an effort to place shutdown cooling Train B in service. Valve SI-405B

is a hydraulic pneumatic operated gate valve in the shutdown cooling return line from

the reactor coolant system to the low-pressure safety injection Pump B suction.

Valve SI-405B has a safety function to open to allow flow through shutdown cooling

Train B. During the test, Valve SI-405B did not stroke fully open within its design time of

15 minutes +/- 225 seconds, and automatically closed per its design. On its second

-10- Enclosure

attempt, Valve SI-405B fully opened, but not within the maximum inservice testing limit

of 370 seconds. Valve SI-405B was declared inoperable and Technical Specification 3.6.3, Containment Isolation Valves, was entered. The licensee closed

valve SI-407B to comply with Technical Specification 3.6.3. This caused shutdown

cooling Train B to become unavailable. Later in Mode 5, Technical Specification 3.6.3

no longer applied and the licensee was allowed to open Valve SI-407B and shutdown

cooling Train B was restored to service. The licensee wrote Condition Report

CR-WF3-2006-3610 to document the open stroke failure of Valve SI-405B. The root

cause analysis determined that the design of SI-405A(B) fluid reservoir is intolerant to

operation with low pressure/level. This condition caused inadequate hydraulic pump

discharge pressure that resulted in the failure of Valve SI-405B to stroke open.

The inspectors noted a similar event on April 17, 2005, that Valve SI-405B failed to fully

open while operations was aligning shutdown cooling system Train B for service in

Mode 4. Operators were performing Operating Procedure OP-903-033 testing to stroke

open Valve SI-405B. During testing, Valve SI-405B failed to reach the full open position

within the maximum allowable time. A root cause evaluation concluded that the valve

was intolerant to minor degradation of various subcomponents in the system, which

caused a diversion of sufficient hydraulic fluid flow to stroke open the valve. The

licensee stated that there was very limited data to support historical failure analysis to

determine the true cause(s) of the failures to operate. Condition

Report CR-WF3-2005-1362 implemented corrective actions to preclude recurrence,

including replacing the hydraulic actuators with a motor-operated actuator in May 2008.

Condition Report CR-WF3-2006-3610 noted that Valve SI-405B had also failed to stroke

open satisfactorily due to various hydraulic system malfunction in October 2003,

April 2002, and October 2000.

Analysis. The deficiency associated with this finding was the failure to establish

corrective measures to prevent recurrence of a significant condition adverse to quality.

Specifically, corrective actions established to address the function of Valve SI-405B to

cycle open were not effectively implemented and failed to prevent recurrence resulting in

Valve SI-405B being declared inoperable. The inspectors determined that the issue was

more than minor in significance since it affected the Mitigating Systems cornerstone

attribute of equipment operability, availability, and reliability of systems that respond to

initiating events. The inspectors evaluated the finding using Inspection

Manual Chapter 0609, Significance Determination Process (SDP), Appendix A, SDP

Phase 1 Screening Worksheet for Initiating Events, Mitigating Systems, and Barrier

cornerstones to assess the safety significance. The finding was determined to be of

very low risk significance because, in each condition identified, it did not represent an

actual loss of a safety function. This finding had crosscutting aspects associated with

the corrective action program component in the problem identification and resolution

area. This assessment was based on the fact that the licensee failed to thoroughly

evaluate the problem such that the resolutions addressed the causes and, therefore,

corrective actions were inadequate to prevent repetition.

Enforcement. 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires

in part, that measures be established to assure that conditions adverse to quality are

promptly identified and corrected. In the case of significant conditions adverse to

-11- Enclosure

quality, the measures shall assure that the cause of the condition is determined and

corrective action taken to preclude repetition. The failure to establish corrective

measures to prevent recurrence of Valve SI-405B failure to stroke open during actual

and test demands conditions impacted the ability of a risk significant system to perform

as designed and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. Corrective

actions for Valve SI-405B per Condition Report CR-WF3-2005-1362 failed to preclude

an additional failure as documented in Condition Report CR-WF3-2006-3610. Because

this violation was of very low safety significance and was entered in the corrective action

program as Condition Report CR-WF3-2006-3610, this violation is being treated as a

noncited violation, consistent with Section VI.A of the NRC Enforcement Policy

(NCV 05000382/2006005-01, Recurring Failure of Valve SI-405B to Open).

1R20 Refueling and Other Outage Activities (71111.20)

The inspectors reviewed the following risk significant refueling items or outage activities

to verify defense in depth commensurate with the outage risk control plan, compliance

with the Technical Specifications, and adherence to commitments in response to

Generic Letter 88-17, Loss of Decay Heat Removal: (1) the risk control plan;

(2) tagging/clearance activities; (3) reactor coolant system instrumentation; (4) electrical

power; (5) decay heat removal; (6) spent fuel pool cooling; (7) inventory control;

(8) reactivity control; (9) containment closure; (10) reduced inventory or midloop

conditions; (11) refueling activities; (12) heatup and cooldown activities; (13) restart

activities; and (14) licensee identification and implementation of appropriate corrective

actions associated with refueling and outage activities. The inspectors containment

inspections included observations of the containment sump for damage and debris; and

supports, braces, and snubbers for evidence of excessive stress, water hammer, or

aging. Documents reviewed by the inspectors included the Refueling Outage 14 Risk

Assessment Plan.

The inspectors completed one sample.

b. Findings

Introduction. A self-revealing violation of very low safety significance of Technical

Specification 6.8.1.a was identified for an inadequate procedure for installing a bolted

joint that provided structural support for the pressurizer. Specifically, the installation

procedure required applying 8750 ft-lbs torque to make up a bolted joint. Following

corrective actions, the licensee discovered that the break away torque on several bolts

exceeded 13,400 ft-lbs. The improper bolt tensioning resulted in failure of 1 of 16 bolts

and the partial cracking of 3 other bolts that potentially could affect the pressurizers

function in a safe shutdown earthquake event. The licensee has since replaced all

pressurizer skirt bolting and installed the bolting to an approved torque specification.

Description. On December 11, 2006, while performing a walkdown of the pressurizer

cubicle, a technician noted that a 2-inch diameter bolt had severed into two pieces. The

bolt was part of a bolted flange that joined the pressurizer skirt to the pressurizer

support structure. The pressurizer skirt is a cylindrical steel shell welded to the bottom

head of the pressurizer and provides structural support to the pressurizer. The

pressurizer is a Category 1 seismic component, which is designed to remain functional

-12- Enclosure

in the event of a safe shutdown earthquake. A steel bolted flange is attached to the

bottom of the pressurizer skirt. The bolted flange on the pressurizer skirt mates with the

top flange of a support structure. The support structure is a weldment anchored by bolts

embedded in the concrete floor below the pressurizer. The failed bolt clamped the

pressurizer skirt flange to the top flange of the support structure. The licensee

ultrasonically test inspected the remaining 15 pressurizer skirt bolts. They discovered

flaws in 3 of the remaining 15 bolts. Licensee replaced all 16 bolts, nuts, and washers.

A licensee review of plant records indicated that an April 1979, Field Change

Procedure FSC-AS-1232, Pressurizer Support Structure, S.I. Tank Support Structures,

R.C. Stops and Supports Structural Steel, installed the bolts to a torque value of

8750 ft-lbs. The procedure did not specify if the bolts should be installed lubricated or

unlubricated, however all the originally installed bolts were lubricated. During

installation, lubrication reduces the friction between thread mating surfaces resulting in a

much higher bolt preload for the same torque values than for an unlubricated bolt.

When replacing the existing bolts, the licensee noted that some bolt break away torque

values exceeded 13,400 ft-lbs. Normally, following a bolted joint installation, some loss

of preload is expected in the range of 5 to 15 percent. Additionally, vibration or

time-varying mechanical or thermal loads on the joint due to system operations can

reduce bolt preload values further. A preliminary root cause evaluation indicates that

the bolts were overloaded during initial preloading at installation. This overloaded

condition reduced the fatigue resistance of the bolting. Thermally induced prying on the

bolts during pressurizer heatups and cooldowns was identified as a potentially

significant low-cycle fatigue load. This loading occurs as a result of temperature

differences between the pressurizer skirt and the support structure. This was an

expected occurrence, but excessive bolt preload increased the magnitude of the prying

and reduce the capacity of the bolts to accept it.

Analysis. The performance deficiency associated with this finding was the failure to

establish appropriate instructions for installing the pressurizer support skirt bolts. The

work instruction did not provide adequate direction to torque the pressurizer skirt bolts

without lubrication. This contributed to the subsequent failure and cracking of several

pressurizer skirt bolts. This finding is more than minor because if left uncorrected it

could have become a more safety significant concern. The finding was associated with

the equipment performance attribute of the Initiating Events cornerstone, and it affected

the cornerstone objective to limit the likelihood of those events that upset plant stability

and challenge critical safety functions during power operations. This finding is of very

low safe significance (Green), because a seismic event was determined to not result in

a loss-of-coolant accident that exceeds the Technical Specification limit for reactor

coolant system leakage. Therefore, this issue screened out in Phase 1 of the Manual

Chapter 0609 Significance Determination Process, because there was no actual loss

of safety funtions.

Enforcement. Technical Specification Section 6.8.1.a requires that written procedures

shall be established, implemented, and maintained covering applicable procedures

recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.

Regulatory Guide 1.33, Appendix A, recommends that maintenance that can affect the

performance of safety-related equipment be performed in accordance with written

procedures, documented instructions, or drawings appropriate to the circumstances.

-13- Enclosure

Contrary to these requirements, the licensee failed to ensure that Field Change

Request FCR-AS-1232 was adequate for the task. The pressurizer skirt bolted work

instruction was inadequate because bolt lubrication was not addressed and resulted in

the failure of 4 of the 16 pressurizer skirt bolts. This issue was entered into the

licensees corrective action program as Condition Report CR-WF3-2006-4274.

(NCV 05000382/2006005-02, Excess Torque Resulting in Pressurizer Skirt Bolt

Failures).

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 four below listed

surveillance activities demonstrated that the structures, systems, and components

tested were capable of performing their intended safety functions. The inspectors either

witnessed or reviewed test data to verify that the following significant surveillance test

attributes were adequate: (1) preconditioning; (2) evaluation of testing impact on the

plant; (3) acceptance criteria; (4) test equipment; (5) procedures; (6) jumper/lifted lead

controls; (7) test data; (8) testing frequency and method demonstrated Technical

Specification operability; (9) test equipment removal; (10) restoration of plant systems;

(11) fulfillment of ASME Code requirements; (12) updating of performance indicator

data; (13) engineering evaluations, root causes, and bases for returning tested

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

correct; (14) reference setting data; and (15) annunciators and alarms setpoints. The

inspectors also verified that the licensee identified and implemented any needed

corrective actions associated with the surveillance testing.

• October 18, 2006; Surveillance Procedure STA-001-005, Leakage Testing of

Air and Nitrogen Accumulators for Safety Related Valves, Revision 7. This test

verified the leakage is within analyzed acceptable limits to fulfill the design

function upon loss of instrument air.

• October 23, 2006; Operations Procedure OP-903-068, Emergency Diesel

Generator, Revision 14. This monthly test verified operability of emergency

diesel Generator A to satisfy Technical Specification requirements.

• November 21, 2006; Maintenance Procedure MM-007-015, Main Steam Safety

Valve Test, Revision 9, Change 0. This test verified the lift pressures for three

of the main steam safety valves prior to entering the refueling outage.

• December 5, 2006; Operations Procedure OP-903-108, Safety Injection Flow

Balance Test, Revision 5, Change 0. This test verified high-pressure safety

injection (HPSI) flow from HPSI Pumps A, B, and AB through HPSI Headers A

and B were within specifications.

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed four samples.

-14- Enclosure

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications (71111.23)

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report, plant drawings,

procedure requirements, and Technical Specifications to ensure that the two below

listed temporary modifications were properly implemented. The inspectors: (1) verified

that the modifications did not have an effect on system operability/availability; (2) verified

that the installation was consistent with modification documents; (3) ensured that the

postinstallation test results were satisfactory and that the impact of the temporary

modifications on permanently installed structures, systems, and components were

supported by the test; (4) verified that the modifications were identified on control room

drawings and that appropriate identification tags were placed on the affected drawings;

and (5) verified that appropriate safety evaluations were completed. The inspectors

verified that licensee identified and implemented any needed corrective actions

associated with temporary modifications.

• December 18, 2006: ER-W3-2006-0264-000, Temporary Air Compressors to

Augment Station Air, a temporary alteration to install a supplemental air

compressor to support Refueling Outage 14.

• December 28, 2006: ER-W3-2006-0375-000 and -001, Install Loose Parts

Monitoring Sensors on the Steam Generators, a temporary alteration to install

acoustic monitoring equipment to monitor for loose parts, associated with the

tube support bar assemblies also known as batwings, in the steam generator

secondary upper shell area.

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed two samples.

b. Findings

No findings of significance were identified.

Cornerstone: Emergency Preparedness (EP)

1EP2 Alert Notification System Testing (71114.02)

a. Inspection Scope

The inspector discussed with licensee staff the status of offsite siren and tone alert radio

systems to determine the adequacy of licensee methods for testing the alert and

notification system in accordance with 10 CFR Part 50, Appendix E. The licensees alert

and notification system testing program was compared with criteria in NUREG-0654,

-15- Enclosure

Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and

Preparedness in Support of Nuclear Power Plants, Revision 1, Federal Emergency

Management Agency (FEMA) Report REP-10, Guide for the Evaluation of Alert and

Notification Systems for Nuclear Power Plants, and the licensees current

FEMA-approved alert and notification system design report, Alert/Notification System,

Waterford-3 Steam Electric Station, dated March 2005. The inspector also reviewed

the following procedures:

• EPP-422, Siren and Helicopter Warning System Maintenance, Revision 3

• EPP-424, Siren Testing and Siren System Administrative Controls, Revision 9

• Emergency Preparedness Desk Guide 16, Siren System Administrative Data,

Revision 11

The inspector completed one sample during the inspection.

b. Findings

No findings of significance were identified.

1EP3 Emergency Response Organization Augmentation Testing (71114.03)

a. Inspection Scope

The inspector discussed with licensee staff the status of primary and backup systems

for augmenting the on-shift emergency response staff to determine the adequacy of

licensee methods for staffing emergency response facilities. The inspector reviewed the

results of 15 licensee augmentation drills performed from August 2005 through

November 2006 as listed in the Attachment to this report, and the listed references

related to the emergency response organization augmentation system, to evaluate the

licensees ability to staff the emergency response facilities in accordance with the

licensee emergency plan and the requirements of 10 CFR Part 50 Appendix E.

The inspector completed one sample during the inspection.

b. Findings

No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)

a. Inspection Scope

The inspector performed an in-office review of Revision 33 to the Waterford Steam

Electric Station Emergency Plan, and Revision 21 to Emergency Plan Implementing

Procedure EP-001-001, Recognition and Classification of Emergency Conditions.

-16- Enclosure

These revisions implemented an NEI 99-01, Methodology for Development of

Emergency Action Levels, Revision 4, emergency action scheme for which prior NRC

approval was obtained.

These revisions were compared to the criteria of NUREG-0654, Criteria for Preparation

and Evaluation of Radiological Emergency Response Plans and Preparedness in

Support of Nuclear Power Plants, Revision 1, to the criteria of NEI 99-01, Methodology

for Development of Emergency Action Levels, Revision 4, to the NRC Safety Analysis

Report dated June 20, 2005, and to the standards in 10 CFR 50.47(b) to determine if

the revisions were adequately conducted according to the requirements of

10 CFR 50.54(q). These reviews were not documented in a safety evaluation report and

did not constitute approval of licensee changes, therefore these revisions are subject to

future inspection.

The inspector completed two samples during the inspection.

b. Findings

No findings of significance were identified.

1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies (71114.05)

a. Inspection Scope

The inspector reviewed the licensees corrective action program requirements in

Procedures EN-LI-102, Corrective Action Process, Revision 2, and EN-LI-119,

Apparent Cause Evaluation Process, Revision 3. The inspector reviewed summaries

of 194 corrective action requests associated with emergency preparedness issues

during calendar years 2005 and 2006 and selected 15 for detailed review against the

program requirements. The inspector reviewed the licensees after-action report for a

significant event (Hurricane Katrina) using the requirements of Procedure UNT-006-10,

Event Notification and Reporting, Revision 17. The inspector evaluated the response

to the corrective action requests to determine the licensees ability to identify, evaluate,

and correct problems in accordance with the licensee program requirements and

10 CFR 50.47(b)(14) and 10 CFR Part 50, Appendix E.

The inspector reviewed the licensees audit program requirements in

Procedure EN-QV-109, Audit Process, Revision 8, quality assurance audits conducted

in 2005 and 2006, and licensee self-assessments of emergency preparedness. The

inspector also reviewed other documents listed in the attachment to this report.

The inspector completed one sample during the inspection.

b. Findings

Introduction: A Green noncited violation was identified for failure to conduct a required

offsite medical drill during calendar year 2005, as required by 10 CFR 50.54(q).

-17- Enclosure

Description: The NRC identified a performance deficiency related to conduct of the

licensees drill and exercise program, in that an annual offsite medical drill involving a

simulated contaminated individual was not performed during calendar year 2005 as

required by the Waterford Steam Electric Station Emergency Plan. The licensee has

two offsite medical support facilities, the Ochsner Clinic and West Jefferson Medical

Center, and typically alternates performing the annual medical drill, so that each facility

has the opportunity to perform the simulated medical treatment of a contaminated

individual. Ochsner Clinic had last performed a drill with simulated medical treatment of

a contaminated individual in 2003, and another drill was scheduled for

October 17, 2006. That drill did not occur.

As a result of the impact of Hurricanes Katrina and Rita on offsite infrastructure,

Ochsner Clinic informally communicated to the licensee that conducting the drill as

scheduled would be a hardship due to reduced staffing at the facilities. On

October 3, 2005, the State of Louisiana Department of Environment Quality wrote to the

U.S. Department of Homeland Security, Federal Emergency Management Agency,

Region VI, requesting postponement of the 2005 Biennial Exercise and several offsite

emergency preparedness drills scheduled for the week beginning October 17, 2005, one

of which being the licensees offsite medical drill. The licensee submitted a request

(ML0529903030) to the NRC on October 24, 2005, requesting a rescheduling

exemption for the 2005 Biennial Exercise. The inspector determined the licensees

request for a scheduling exemption was limited to the 2005 Biennial exercise and did not

include a request for relief from the emergency plan requirement to conduct the offsite

medical drill.

The U.S. Department of Homeland Security, Federal Emergency Management Agency,

Region VI, responded on October 14, 2005, to the State of Louisiana letter of

October 3, 2005, approving the postponement of the 2005 Biennial Exercise and the

offsite emergency preparedness drills scheduled for the week beginning

October 17, 2005. The Federal Emergency Management Agency, Region VI, letter

stated in part, . . . we concur that . . .the out of sequence drills scheduled for the week

of October 17th should be postponed. . . ., and . . . we will be contacting your

organization in the near future to reschedule the drills. . . . The inspector determined

that the offsite medical drill was not subsequently rescheduled. The NRC responded

(ML053270770) on November 17, 2005, to the licensees letter of October 24, 2005,

approving rescheduling the 2005 Biennial Exercise to 2006; this exercise was

subsequently conducted June 28, 2006.

The inspector determined the licensee conducted an offsite medical drill with West

Jefferson Medical Center on October 25, 2006 (Drill 2006-09), which the licensee

intended as meeting the 2006 annual drill requirement, not the rescheduled 2005 drill

requirement. West Jefferson Medical Center had also drilled the simulated medical

treatment of a contaminated individual in 2004. Ochsner Clinic is scheduled to perform

a drill with simulated medical treatment of a contaminated individual in October 2007,

approximately 4 years after their most recent medical drill in 2003.

Analysis: The inspector determined that the failure to conduct a required offsite medical

drill is a performance deficiency because the licensee failed to meet a requirement of

the Waterford Steam Electric Station Emergency Plan, and the cause was within the

-18- Enclosure

licenseess ability to foresee, correct, and prevent. The finding had a credible impact on

the Emergency Preparedness cornerstone objective because it involved the ability to

maintain the proficiency of offsite personnel whose assistance may be needed in the

event of a radiation emergency, and affected the attributes of emergency response

organization readiness and performance, and offsite emergency preparedness. This

finding is more than minor because a degraded proficiency in providing appropriate

medical treatment for a contaminated individual has a potential impact on the safety of

licensee employees and the public. This finding was evaluated using the Emergency

Preparedness Significance Determination Process and was determined to be of very low

safety significance because the licensee failed to conduct only one required drill during

the inspection period January 2005 through December 2006, and the drill was not

appropriately rescheduled with NRC approval.

Enforcement: Part 50.54(q) of Title 10 of the Code of Federal Regulations states, in

part, A licensee authorized to possess and operate a nuclear power reactor shall follow

and maintain in effect emergency plans which meet the standards in §50.47(b) and the

requirements in Appendix E of this part. 10 CFR 50.47(b)(14) states, in part, ...periodic

drills will be conducted to develop and maintain key skills... 10 CFR Part 50,

Appendix E, IV, F.1 states, in part, The program to provide for: (a) The training of

employees and exercising, by periodic drills, of radiation emergency plans to ensure that

employees of the licensee are familiar with their specific emergency response duties,

and (b) The participation in the training and drills by other persons whose assistance

may be needed in the event of a radiation emergency shall be described. . . vii. Medical

Support Personnel. Section 8.1.2.4(4) of the Waterford Steam Electric Station

Emergency Plan, Revision 33, states, in part, A medical emergency drill, involving a

simulated contaminated individual, which includes provisions for participation by the

local support services (i.e. ambulance and offsite medical treatment facility) shall be

conducted annually. Between January 1 and December 31, 2005, the licensee did not

conduct a drill involving a simulated contaminated individual with provision for

participation by local support services. Because this failure is of very low safety

significance and has been entered into the licensees corrective action system

(Condition Report 2006-4429), this violation is being treated as an noncited violation

consistent with Section VI.A of the NRC Enforcement Policy:

NCV 05000382/2006-005-03, Failure to Conduct a Required Offsite Medical Drill in

2005.

2. RADIATION SAFETY

Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess the licensees performance in implementing physical

and administrative controls for airborne radioactivity areas, radiation areas, high

radiation areas, and worker adherence to these controls. The inspector used the

requirements in 10 CFR Part 20, the Technical Specifications, and the licensees

procedures required by Technical Specifications as criteria for determining compliance.

-19- Enclosure

During the inspection, the inspector interviewed the radiation protection manager,

radiation protection supervisors, and radiation workers. The inspector performed

independent radiation dose rate measurements and reviewed the following items:

• Performance indicator events and associated documentation packages reported

by the licensee in the Occupational Radiation Safety Cornerstone

• Controls (surveys, posting, and barricades) of multiple radiation, high radiation,

or airborne radioactivity areas

• Radiation work permits, procedures, engineering controls, and air sampler

locations

• Conformity of electronic personal dosimeter alarm set points with survey

indications and plant policy; workers knowledge of required actions when their

electronic personnel dosimeter noticeably malfunctions or alarms

• Barrier integrity and performance of engineering controls in three airborne

radioactivity areas

• Adequacy of the licensees internal dose assessment for any actual internal

exposure greater than 50 millirem Committed Effective Dose Equivalent

• Physical and programmatic controls for highly activated or contaminated

materials (nonfuel) stored within spent fuel and other storage pools

• Self-assessments, audits, and special reports related to the access control

program since the last inspection

• Corrective action documents related to access controls

• Licensee actions in cases of repetitive deficiencies or significant individual

deficiencies

• Radiation work permit briefings and worker instructions

• Adequacy of radiological controls such as, required surveys, radiation protection

job coverage, and contamination controls during job performance

• Dosimetry placement in high radiation work areas with significant dose rate

gradients

• Changes in licensee procedural controls of high dose rate - high radiation areas

and very high radiation areas

• Controls for special areas that have the potential to become very high radiation

areas during certain plant operations

• Posting and locking of entrances to all accessible high dose rate - high radiation

areas and very high radiation areas

-20- Enclosure

• Radiation worker and radiation protection technician performance with respect to

radiation protection work requirements

The inspector completed 17 of the required 21 samples.

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual and

collective radiation exposures as low as is reasonably achievable (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 ALARA procedures

• Interfaces between operations, radiation protection, maintenance, maintenance

planning, scheduling, and engineering groups

• Integration of ALARA requirements into work procedure and radiation work permit

documents

• Exposure tracking system

• Workers use of the low dose waiting areas

• First-line job supervisors contribution to ensuring work activities are conducted in

a dose efficient manner

• Source-term control strategy or justifications for not pursuing such exposure

reduction initiatives

• Specific sources identified by the licensee for exposure reduction actions and

priorities established for these actions, and results since the last refueling cycle

• Radiation worker and radiation protection technician performance during work

activities in radiation areas, airborne radioactivity areas, or high radiation areas

The inspector completed 4 of the required 15 samples and 7 of the optional samples.

-21- Enclosure

b. Findings

No findings of significance were identified.

4. OTHER ACTIVITIES (OA)

4OA1 Performance Indicator Verification (71151)

.1 Mitigating Systems and Barrier Integrity

a. Inspection Scope

The inspectors sampled licensee submittals for the three performance indicators listed

below for the period from July 1, 2004, through September 30, 2006. The definitions and

guidance of Nuclear Energy Institute 99-02, Regulatory Assessment Indicator

Guideline, Revision 4, were used to verify the licensees basis for reporting each data

element in order to verify the accuracy of performance indicator data reported during the

assessment period. The inspectors reviewed licensee event reports, monthly operating

reports, and operating logs as part of the assessment. Licensee performance indicator

data were also reviewed against the requirements of Procedure EN-LI-114,

Performance Indicator Process, Revision 1.

• Safety System Functional Failures

• Reactor Coolant System Specific Activity

• Reactor Coolant System Leakage

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed three samples.

b. Findings

No findings of significance were identified.

.2 (Closed) Temporary Instruction 2515/169: Mitigating Systems Performance Index

Verification

a. Inspection Scope

During this inspection period, the inspectors completed a review of the licensees

implementation of the mitigating systems performance index in accordance with the

guidance provided in Temporary Instruction 2515/169. The review examined the

licensees mitigating systems performance index basis documents (W3-SA-06-00001,

Revision 0) and verified the established system boundaries and monitored components

were consistent with guidance provided in NEI 99-02, Reactor Oversight Process

Performance Indicators, Revision 4. The inspectors verified that the licensee did not

include credit for unavailability hours for short term unavailability or operator recovery

actions to restore the risk-significant function as is allowed by NEI 99-02.

Additionally, the inspectors reviewed the baseline mitigating systems performance index

unavailability time using plant specific values for the period of 2002 to 2004. The

verification included all planned and unplanned unavailability. The plant specific data for

-22- Enclosure

2005 to 2006 was also reviewed to ensure the licensee properly accounted for the actual

unavailability hours of mitigating systems performance index systems. For the same

period, the mitigating systems performance index component unreliability data was

examined to ensure the licensee identified all failures of monitored components. The

accuracy and completeness of the reported unavailability and unreliability data was

verified by reviewing operating logs, condition reports, and work order documents. The

unavailability and unreliability data was compared with performance indicator data

submitted to the NRC to ensure that any discrepancies would not result in a change to

the index color.

b. Findings

No findings of significance were identified. This completes the inspection requirements

for this Temporary Instruction.

.3 Emergency Preparedness

a. Inspection Scope

The inspector reviewed licensee evaluations for the three Emergency Preparedness

cornerstone performance indicators of Drill and Exercise Performance, Emergency

Response Organization Participation, and Alert and Notification System Reliability, for the

period October 1, 2005, through September 30, 2006. The definitions and guidance of

NEI 99-02, Regulatory Assessment Indicator Guideline, Revisions 2 through 4, and the

licensee Performance Indicator Procedures EN-LI-114, Performance Indicator Process,

Revision 2, and EN-EP-201, Emergency Planning Performance Indicators, Revision 5,

were used to verify the accuracy of the licensees evaluations for each performance

indicator reported during the assessment period.

The inspector reviewed 100 percent of drill and exercise scenarios and licensed operator

simulator training sessions, notification forms, and attendance and critique records

associated with training sessions, drills, and exercises conducted during the verification

period. The inspector reviewed 18 emergency responder qualification, training, and drill

participation records. The inspector reviewed alert and notification system testing

procedures, maintenance records, and 100 percent of siren test records. The inspector

also reviewed other documents listed in the attachment to this report.

The inspector completed three samples during the inspection.

b. Findings

No findings of significance were identified.

.4 Occupational Radiation Safety and Public Radiation Safety

a. Inspection Scope

The inspector sampled licensee submittals for the performance indicators listed below for

the period from April 2006 through October 2006. To verify the accuracy of the

performance indicator data reported during that period, performance indicator definitions

-23- Enclosure

and guidance contained in NEI 99-02, Regulatory Assessment Indicator Guideline,

Revision 2, were used to verify the basis in reporting for each data element.

• Occupational Exposure Control Effectiveness Performance Indicators

Licensee records reviewed included corrective action documentation that identified

occurrences in high radiation areas with dose rates greater than 1,000 millirem per hour

at 30 centimeters (as defined in technical specifications), very high radiation areas (as

defined in 10 CFR 20.1003), and unplanned personnel exposures (as defined in

NEI 99-02). Additional records reviewed included ALARA records and whole-body

counts of selected individual exposures. The inspector interviewed licensee personnel

that were accountable for collecting and evaluating the performance indicator data. In

addition, the inspector toured plant areas to verify that high radiation and very high

radiation areas were properly controlled.

The inspector completed the required sample (1) in this cornerstone.

• Radiological Effluent Technical Specification/Offsite Dose Calculation Manual

Radiological Effluent Occurrences

Licensee records reviewed included corrective action documentation that identified

occurrences for liquid or gaseous effluent releases that exceeded performance indicator

thresholds and those reported to the NRC. The inspector interviewed licensee personnel

that were accountable for collecting and evaluating the performance indicator data.

The inspector completed the required sample (1) in this cornerstone.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152)

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

The inspectors performed a daily screening of items entered into the licensees corrective

action program. This assessment was accomplished by reviewing condition reports and

event trend reports and attending daily operational meetings. The inspectors:

(1) verified that equipment, human performance, and program issues were being

identified by the licensee at an appropriate threshold and that the issues were entered

into the corrective action program; (2) verified that corrective actions were commensurate

with the significance of the issue; and (3) identified conditions that might warrant

additional follow-up through other baseline inspection procedures.

b. Findings

No findings of significance were identified.

-24- Enclosure

.2 Selected Issue Follow-up Inspection

a. Inspection Scope

In addition to the routine review, the inspectors selected the one below listed issue for a

more in-depth review. The inspectors considered the following during the review of the

licensees actions: (1) complete and accurate identification of the problem in a timely

manner; (2) evaluation and disposition of operability/reportability issues; (3) consideration

of extent of condition, generic implications, common cause, and previous occurrences;

(4) classification and prioritization of the resolution of the problem; (5) identification of

root and contributing causes of the problem; (6) identification of corrective actions; and

(7) completion of corrective actions in a timely manner.

• December 18, 2006: Operator Workarounds

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed one sample.

b. Findings

No findings of significance were identified.

.3 Semiannual Trend Review

a. Inspection Scope

The inspectors completed a semiannual trend review of repetitive or closely related

issues associated with the essential chiller low evaporator pressure trips that were

documented in condition reports, system and component health reports, quality

assurance audits, trend reports, the licensees internal performance indicators, and NRC

inspection reports to identify trends that might indicate the existence of more safety

significant issues. The inspectors review consisted of the 6-month period of July 1 to

December 31, 2006. When warranted, some of the samples expanded beyond those

dates to fully assess the issue. The inspectors also reviewed corrective action program

items associated with troubleshooting. The inspectors compared and contrasted their

results with the results contained in the licensees quarterly trend reports. Corrective

actions associated with a sample of the issues identified in the licensees trend report

were reviewed for adequacy.

Documents reviewed by the inspectors are listed in the attachment. The inspectors

completed one sample.

b. Findings

No findings of significance were identified.

-25- Enclosure

.4 Annual Sample Review

a. Inspection Scope

The emergency preparedness inspector selected 15 condition reports for detailed review.

The reports were reviewed to ensure that the full extent of the issues were identified, an

appropriate evaluation was performed, and appropriate corrective actions were specified

and prioritized. The inspector evaluated the condition reports using licensee

Procedures EN-LI-102, Corrective Action Process, Revision 2, and EN-LI-119,

Apparent Cause Evaluation Process, Revision 3.

The health physics inspector evaluated the effectiveness of the licensees problem

identification and resolution process with respect to the following inspection areas:

• Access Control to Radiologically Significant Areas (Section 2OS1)

• ALARA Planning and Controls (Section 2OS2)

b. Findings and Observations

No findings of significance were identified.

4OA3 Event Follow-up (71153)

.1 (Closed) LER 05000382/2005-005-00: Manual Reactor Trip Upon Loss of All Circulating

Water Pumps and Lowering Condenser Vacuum

On November 11, 2005, the licensee manually tripped the reactor due to lowering main

condenser vacuum, caused by a loss of all circulating water pumps. Lowering main

condenser vacuum resulted in loss of main feedwater to the steam generators causing

steam generator levels to lower resulting in an automatic actuation of the emergency

feedwater system to restore steam generator level. Failure mode analysis identified a

degraded timer relay in the CW pump discharge valve control circuit as the most likely

cause. The relay was replaced prior to plant start up. The LER was reviewed by the

inspectors and no findings of significance were identified and no violation of NRC

requirements occurred. The licensee documented the failed equipment in Condition

Report CR-WF3-2005-4593. This LER is closed.

.2 (Closed) LER 05000382/2005-002-01: RCS Leakage Detection Instrumentation and

Regulatory Guide 1.45

On November 15, 2005, the licensee determined that Technical Specification 3.4.5.1,

Leakage Detection Systems, did not meet the design requirements of Regulatory

Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection System.

Specifically, the containment fan cooler condensate flow switches did not meet the

design requirements for detecting a one gallon per minute reactor coolant system leak.

This deficiency was previously dispositioned in NRC Inspection

Report 05000382/2005005, Section 1R22, Surveillance Testing, as a Green noncited

violation (NCV 05000382/2005005-01). This LER is closed.

-26- Enclosure

4OA5 Other Activities

.1 Institute of Nuclear Power Operations (INPO) Audit and Evaluation Review

The inspectors completed a review of the INPO audit and evaluation report for Entergy

Operations Waterford 3 Steam Electric Station during this inspection period. The INPO

audit and evaluation was performed during spring of 2006.

.2 (Closed) NRC Temporary Instruction 2515/166: PWR Containment Sump Blockage

a. Inspection Scope

The inspectors reviewed Waterford 3's implementation of plant modifications and

procedure changes committed to in their response to Generic Letter 2004-02.

The inspectors observed installation of the containment recirculation sump strainers and

relocation of tri-sodium phosphate baskets. In addition, the inspectors verified that

Waterford 3 has implemented specific procedure changes to control tags, labels, tape,

and other objects inside the containment building. At the time of the exit meeting,

Waterford 3 was in the final stages of implementing changes to the containment coatings

assessment program, the latent debris assessment program, and the containment

strainer inspection program.

At the time of the inspection, industry testing for chemical effects on containment

recirculation sumps was not complete. Since the testing was not complete, Waterford 3

evaluated the new recirculation sump modifications to the original design basis,

Regulatory Guide 1.82, Revision 0. The inspectors reviewed the 10CFR 50.59

evaluation to verify that the design meets the original design basis.

b. Findings

No findings of significance were identified. This completes the inspection requirements

for this Temporary Instruction.

4OA6 Meetings, Including Exit

Exit Meeting Summary

.1 On September 25, 2006, the operations inspector discussed the inspection results of the

licensed operator annual requalification examination with Mr. A. Hill, Operations Training

Supervisor. A telephone exit was held with Mr. Hill on September 25, 2006. The

licensee acknowledged the findings presented in both the briefing and the final exit

meeting.

.2 On December 1, 2006, the health physics inspector presented the Occupational

Radiation Safety inspection results to Ms. K. Cook, Acting General Manager, Plant

Operations, and other members of the staff who acknowledged the findings. The

inspector confirmed that proprietary information was not provided or examined during the

inspection.

-27- Enclosure

.3 On December 8, 2006, the inspectors presented the results of Temporary

Instruction 2515/166 to Mr. K. Walsh, Waterford 3 Site Vice President, and other

members of licensee management. Licensee management acknowledged the inspection

findings. The inspectors identified that they had reviewed proprietary information but had

returned it to licensee personnel.

.4 On December 20, 2006, the inspector presented the inspection results to Mr. K. Walsh,

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

inspector confirmed that proprietary information was not provided or examined during the

inspection.

.5 On January 16, 2007, the resident inspectors presented the inspection results to

Mr. K. Walsh and other members of licensee management at the conclusion of the

inspection. The licensee acknowledged the findings presented. The inspectors asked

the licensee whether any materials examined during the inspection should be considered

proprietary. No proprietary information was identified.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by the

licensee and is a violation of NRC requirements, which meets the criteria of Section VI of

the NRC Enforcement Policy for being dispositioned as an NCV.

• Part 50 of Title 10 of the Code of Federal Regulations, Appendix E, IV, B,

requires a licensee establish emergency action levels based on in-plant

conditions and instrumentation. Contrary to this, the licensee changed

radiological accident assumptions in their Updated Final Safety Analysis Report in

1994, 2001, and 2003, and corresponding changes to emergency action levels

were not made. This was identified in the licensees corrective action program as

Condition Report 2005-3292. This finding is of very low safety significance

because the affected emergency action levels were at the Notification of Unusual

Event and Alert emergency classifications and did not affect classification at the

Site Area Emergency or General Emergency levels.

ATTACHMENT: SUPPLEMENTAL INFORMATION

-28- Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

S. Anders, Superintendent, Plant Security

J. Brawley, ALARA Coordinator, Radiation Protection

K. Cook, Acting General manager, Plant Operations

L. Dauzat, Operations Supervisor, Radiation Protection

R. Dodds, Manager, Operations

C. Fugate, Assistant Manager, Operations (Shift)

T. Gaudet, Manager, Quality Assurance

J. Lewis, Manager, Emergency Preparedness

C. Miller, Supervisor, Radiation Protection

R. Murillo, Manager, Licensing

R. Peters, Director, Planning and Scheduling

B. Pilutti, Manager, Radiation Protection

O. Pipkins, Senior Licensing Engineer

R. Putnam, Manager, Engineering Programs

G. Scott, Licensing Engineer

K. Walsh, Vice President, Operations

B. Williams, Director, Engineering

R. Williams, Licensing Engineer

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000382/2006-005-01 NCV Recurring Failure of Valve SI-405B to Open

(Section 1R19)

05000382/2006-005-02 NCV Excess Torque Resulting in Pressurizer Skirt Bolt

Failures (Section 1R20)

05000382/2006-005-03 NCV Failure to Conduct a Required Offsite medical Drill in

2005 (Section 1EP5)

Closed

05000382/2006-005-01 NCV Recurring Failure of Valve SI-405B to Open

(Section 1R19)

05000382/2006-005-02 NCV Excess Torque Resulting in Pressurizer Skirt Bolt

Failures (Section 1R20)

05000382/2005-002-01 LER RCS Leakage Detection Instrumentation and Regulatory

Guide 1.45

05000382/2005-005-00 LER Manual Reactor Trip Upon Loss of All Circulating Water

Pumps and Lowering Condenser Vacuum

05000382/2006-005-03 NCV Failure to Conduct a Required Offsite medical Drill in

2005 (Section 1EP5)

A-1 Attachment

LIST OF DOCUMENTS REVIEWED

Section 1R04: Equipment Alignment (71111.04)

Procedures

Number Title Revision

OP-002-005 Chemical and Volume Control Revision 21

Miscellaneous Documents

Updated Final Safety Analysis Report

Flow Diagram - Chemical and Volume Control System, G168, Sheet 2, Rev. 48

Section 1R05: Fire Protection (71111.05)

Procedure

NUMBER TITLE REVISION

Administrative Procedure Fire Protection Program 9

UNT-005-013

Operating Procedure 009-004 Fire Protection 11-8

Maintenance Procedure MM- Fire Extinguisher Inspection and 13

007-010 Extinguisher Replacement

Administrative Fire Protection Program 9

Procedure UNT-005-013

Fire Protection Procedure FP- Fire Protection System 17

001-015 Impairments

Fire Protection Procedure FP- Transient Combustibles 19

001-017

Training Manual Procedure Fire Protection Training 11-4

NTP-202

Section 1R12: Maintenance Rule (71111.12)

Procedures

Number Title Revision

DC-121 Maintenance Rule 1

NUMARC 93-01 Industry Guideline for Monitoring the Effectiveness 3

of Maintenance at Nuclear Power Plants

A-2 Attachment

Condition Reports

CR-WF3-2005-3084 CR-WF3-2006-0395 CR-WF3-2006-1926

CR-WF3-2005-3270 CR-WF3-2006-0397 CR-WF3-2006-2247

CR-WF3-2005-3692 CR-WF3-2006-0467 CR-WF3-2006-2384

CR-WF3-2005-4852 CR-WF3-2006-0733 CR-WF3-2006-2398

CR-WF3-2006-0007 CR-WF3-2006-0796 CR-WF3-2006-2736

CR-WF3-2006-0191 CR-WF3-2006-0857 CR-WF3-2006-2819

CR-WF3-2006-0212 CR-WF3-2006-1266 CR-WF3-2006-3276

CR-WF3-2006-0346 CR-WF3-2006-1297 CR-WF3-2006-3285

Miscellaneous Documents

Engineering Report W-SE- Waterford 3 Maintenance Rule Periodic (a)(3) 0

2005-001 Assessment

Section 1R15: Operability Evaluations

Procedures:

NUMBER TITLE REVISION

EN-OP-104 Operability Evaluation 1

OP-903-097 Pressurizer Heater Capacity Verification 8

OP-035-000 Notification Matrix 6

Miscellaneous Documents

NUMBER TITLE/SUBJECT REVISION

Waterford 3 LER Loose Breaker Fuse Rendered One Bank of Pressurizer 0

2003-001-000 Proportional Heaters Inoperable

Condition Reports

CR-WF3-2006-3125 CR-WF3-2006-4501 CR-WF3-1997-2226

CR-WF3-2006-3128 CR-WF3-1997-2491 CR-WF3-2002-1757

CR-WF3-2004-0846 CR-WF3-2006-3357 CR-WF3-2003-0827

CR-WF3-2006-4510 CR-WF3-2006-2418

CR-WF3-2000-0228

Section 1R19: Postmaintenance Testing (71111.19)

A-3 Attachment

Procedures

NUMBER TITLE REVISION

MI-003-115 Startup and Control Drawer Calibration Channel 3

1 or 2

MI-012-012 Removal and Installation of Excore Detectors 3

OP-903-101 Startup Channel Functional Test Channel 1 and 6

2

OP-903-033 Cold Shutdown IST Valve Tests 20

OP-903-003 Charging Pump operability Check Rev. 11.

Change 1

ENS-MA-114 Post Maintenance Testing 5

Miscellaneous Documents

NUMBER TITLE/SUBJECT REVISION

CEP-IST-1 IST Bases Document 3

Condition Reports

CR-WF3-2006-3610 CR-WF3-2000-1347

CR-WF3-2005-1362

Work Orders

44779-01, 65817-01, 73483,

Procedures

Number Title Revision

OP-903-003 Charging Pump Operability Check Revision 11,

Change 1

OP-903-121 Safety Systems Quarterly IST Valve Tests 7

OP-903-127 Reactor Trip Circuit Breaker Post 3

Maintenance Test

OP-903-013 Monthly Channel Checks 14

OP-903-011 High Pressure Safety Injection Pump 9

Preservice Operability Check

A-4 Attachment

CEP-IST-1 IST Bases Document 3

Work Orders

50947-1, 51041014-01, 63598-01,

Section 1R23: Temporary Plant Modifications (71111.23)

Procedure

NUMBER TITLE REVISIONS

EN-LI-113 Licensing Basis Document Change 1

Process

Miscellaneous Documents

NUMBER TITLE/SUBJECT REVISION

ER-W3-2006-0264- Temporary Air Compressors to Augment Station Air 0

000

NRC Information Loose Part Detection and Computerized Eddy Current 0

Notice 2004-17 Data Analysis in Steam Generators

ER-W3-2006-0375- Install Loose Parts Monitoring Sensors on the Steam 0

000 Generators

Condition Reports

CR-WF3-2006-4524

Work Order 84381

Section 1EP3: Emergency Response Organization Augmentation Testing (71114.03)

EP-002-015, Emergency Responder Activation, Revision 8

EP-003-070, Emergency Communications Systems Routine Testing, Revision 24

EPP-462, Evaluation of Pager Tests, Revision 0

Desk Guide 20, Evaluation of Pager Tests, Revision 20

Drill 2005-03, Unannounced Off-Hours Callout Drill, conducted August 11, 2005

Drill 2005-07, Backup Emergency Response Organization Pager Code Drill, conducted

A-5 Attachment

September 12, 2005

Drill 2006-07, Backup Emergency Response Organization Pager Code Drill, conducted

November 14, 2006

Evaluation Worksheets for Pager Tests conducted: January 31, February 22, March 18, April 9,

May 9, June 22, July 25, November 8 (all 2005), January 15, February 22, September 6, and

November 2 (all 2006)

Section 1EP4: Emergency Action Level and Emergency Plan Changes (71114.04)

Safety Evaluation Report, Proposed Emergency Action Levels Based on Revision 4 to Nuclear

Energy Institute 99-01, Entergy Operations Inc., Waterford Steam Electric Station, Unit 3, dated

June 20, 2005

Section 1EP5: Correction of Emergency Preparedness Weaknesses and

Deficiencies (71114.05)

EP-002-150, Emergency Plan Implementing Records,

W3D3-2005-011, Hurricane Katrina Event Report

QA-7-2005-WF3-1, Quality Assurance Audit Report: Emergency Plan

QA-7-2005-WF3-1, Followup to the 2005 QA Emergency Planning Audit

QA-7-2006-WF3-1, Quality Assurance Audit Report: Emergency Plan

QS-2005-W3-003, Quality Assurance Surveillance Report: Emergency Plan Respiratory

Equipment and Reviews of CR-ECH-2004-00096 and CR-ECH-2004-0389"

Evaluation Reports for:

Drill 2005-01, conducted February 17, 2005

Drill 2005-04, conducted August 4, 2005

Drill 2005-06, conducted December 7, 2005

Drill 2006-02, conducted May 25, 2006

Drill 2006-03, conducted June 28, 2006

Drill 2006-04, conducted July 27, 2006

Drill 2006-05, conducted September 21, 2006

Drill 2006-09, Offsite Medical Response Drill, conducted October 25, 2006

LO-WLO-2005-0043, 1st and 2nd Quarter 2005 Roll-Up Assessment: Emergency Planning

Department

LO-WLO-2005-0082, 3rd Quarter 2005 Roll-Up Assessment: Emergency Planning Department

A-6 Attachment

LO-WLO-2005-0106, 4th Quarter 2005 Roll-Up Assessment: Emergency Planning Department

LO-WLO-2006-0044, 1st Quarter 2006 Roll-Up Assessment: Emergency Planning Department

LO-WLO-2006-0069, 2nd Quarter 2006 Roll-Up Assessment: Emergency Planning Department

LO-WLO-2006-0103, 3rd Quarter 2006 Roll-Up Assessment: Emergency Planning Department

SNAPSHOT Assessment, Siren Battery Storage, October 19, 2005

Assessment Report LO-WLO-2005, Emergency Planning Performance Indicator Assessment,

October 26, 2005

Assessment Report LO-WLO-2006-00, Emergency Response Organization Staffing, September

25, 2006

Assessment Report LO-WLO-2006-041, Emergency Planning Performance Indicator

Assessment, April 21, 2006

Assessment Report LO-WLO-2006-000, Emergency Planning Performance Indicator

Assessment, October 29, 2006

EN-HU-103, Human Performance Error Reviews, Revision 0

Condition Reports:

2005-0046, -1060, -3292, -3471, -3602, -3702, -4407, -4899

2006-0234, -897, -1195, -1639, -1900, -2088, -2852, -3247, -4418

Section 2OS1: Access Controls to Radiologically Significant Areas (71121.01)

Corrective Action Documents

2006-03645, 2006-03661, 2006-03721,2006-03736, 2006-03739, 2006-03741, 2006-03812,

Procedures

ENS-RP-102, Radiological Control, Revision 0

ENS-RP-105, Radiation Work Permits, Revision 7

ENS-RP-106, Radiological Survey Documentation, Revision 0

Audits and Assessments

LO-WLO-2006-00067-01, Access Control to Radiologically Significant Areas

Radiation Work Permit

RWP-2006-0005, Tours and Inspections

RWP-2006-0508, Reactor Coolant Pump Motor and Seal Replacement

RWP-2006-0509, Primary Manways

RWP-2006-0510, Nozzle Dams

RWP-2006-0511, Eddy Current Steam Generators

RWP-2006-0600, Health Physics Surveys

RWP-2006-0601, Rigging RWP

A-7 Attachment

RWP-2006-0603, Minor Maintenance Locked High Radiation Areas

RWP-2006-0608, Safety Injection Sump Installation

RWP-2006-0614, Pressurizer Manway and Valves

RWP-2006-0618, Remove/Replace Insulation in the Reactor Building and Annulus

RWP-2006-0702, Reactor Disassembly

RWP-2006-0708, Remove/Replace Startup Detector No. 2

Section 2OS2: ALARA Planning and Controls (71121.02)

Procedures

CE-002-006, Maintaining Reactor Coolant Chemistry, Revision 13

EN-RP-104, Personnel Contamination Events, Revision 1

EN-RP-109, Hot Spot Program, Revision 2

EN-RP-110, ALARA Program, Revision 2

HP-002-222, Steam Generator Radiological Controls, Revision 7

Section 4OA1: Performance Indicator Verification (71151)

Procedures

EN-LI-114, Performance Indicator Process, Revision 1

Miscellaneous Documents

QA/Oversight Observations, November 28, 2008

RF-14 Actual RCS Cleanup, November 28, 2008

4OA2 Identification and Resolution of Problems (71152)

Procedure

NUMBER TITLE REVISIONS

OP-903-094 ESTAS Subgroup Relay Test - Operating 10

EN-LI-113 Licensing Basis Document Change 1

Process

EN-LI-102 Corrective Action Process 7

OP-002-004 Chilled Water System 12

OP-002-003 Component Cooling Water System 12 and 13

EP-001-001 Recognition and Classification of 20-2 and 21

Emergency Conditions

EP-002-010 Notifications and Communications 30

EP-002-052 Protective Action Guidelines 19

A-8 Attachment

Miscellaneous Documents

NUMBER TITLE/SUBJECT REVISION

ER-W3-00-0541-00 Evaluate the Essential Chilled Water Leaving 0

Temperature Setpoint

Quarterly Trend 2nd Quarter 2006 0

Report

Desk Guide 17 Drill Control Team Documentation 2

Training Evaluation Action Request 2006-1230

Waterford Steam Electric Station Emergency Plan 33

Operations Department Performance Indicator 2006

Condition Reports

CR-WF3-1993-0265 CR-WF3-1996-1852 CR-WF3-2002-1876

CR-WF3-1993-0289 CR-WF3-1997-0028 CR-WF3-2006-3402

CR-WF3-1994-0642 CR-WF3-1997-0288 CR-WF3-2006-3487

CR-WF3-1995-0963 CR-WF3-1997-2778 CR-WF3-2006-4165

CR-WF3-1995-0963 CR-WF3-1999-0816 CR-WF3-2006-0609

CR-WF3-1995-1047 CR-WF3-2000-0054 CR-WF3-2006-1145

CR-WF3-1996-0043 CR-WF3-2000-0150 CR-WF3-2006-3402

CR-WF3-1996-0084 CR-WF3-2000-1553

4OA5 Other: NRC Temporary Instructions 2515/166

Calculation

NUMBER TITLE REVISION

GENE-0000-0054-9349 SIS Sump Strainer Stress Report 0

Engineering Requests

NUMBER TITLE REVISION

ER-W3-2003-0394-001 Safety Injection Sump 0

Modifications

Drawings

NUMBER TITLE REVISION

06-594, Sht. #1 Waterford 3 Containment & A

Strainers06-595, Sht. #2 Waterford 3 Containment & A

Strainers06-596, Sht. #3 Waterford 3 Containment & A

A-9 Attachment

Strainers06-597, Sht. #4 Waterford 3 Containment & A

Strainers06-598, Sht. #5 Waterford 3 Containment & A

Strainers06-486, Sht. #0 LT-SI-7145AS & LT-SI-7145BS 4

Local Mounts

B430, Sht. X-23J-45 Instrument Installation Details 1

B430, Sht. X-23D-8A Instrument Installation Details 2

B430, Sht. X-23J-25A Instrument Installation Details 5

B430, Sht. X-23J-44 Instrument Installation Details 2

B430, Sht. X-23J-28 Instrument Installation Details 8

Condition Reports

CR-WF3-2006-03273

Miscellaneous

NUMBER TITLE REVISION/DATE

LPL-EQMI-08.01 Environmental Qualification 9

Maintenance Input for Rosemount

Model 1153 Series A, B &D, 1154 &

1154 Series H Transmitters and

1159 Remote Seals

SQ-IC-03 Rosemount Pressure Transmitters 13

LPL-EQA-08.01B Environmental Qualification 3

Assessment for the Rosemount

1154 Transmitters Used in the

Waterford SES Unit No. 3

LPL-EQA-08.01F Environmental Qualification 0

Assessment for the Rosemount

1159 Remote Seal System Used in

the Waterford SES Unit No. 3

NOECP-333, Att. 7.1 Construction Material Testing 2

Document, Grout/Mortar Data, Lab

No. W2576, WO No. 76399

A-10 Attachment