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IR 05000382-03-004, on 12/29/2002-03/22/2003; Entergy Operations, Inc.; Waterford Steam Electric Station; Unit 3; Event Followup
	ML031070497
Person / Time
Site:	Waterford
Issue date:	04/17/2003
From:	William Jones; NRC/RGN-IV/DRP/RPB-E
To:	Venable J; Entergy Operations
References
	IR-03-004
	Download: ML031070497 (24)
	v • d • e

April 17, 2003

Joseph E. Venable

Vice President Operations

Waterford 3

Entergy Operations, Inc.

17265 River Road

Killona, Louisiana 70066-0751

SUBJECT: NRC INSPECTION REPORT 50-382/03-04

Dear Mr. Venable:

On March 22, 2003, the NRC completed an inspection at your Waterford Steam Electric

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

on March 24, 2003, with you and other members of your staff.

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

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

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

personnel.

Based on the results of this inspection, one self-revealing finding was identified that was

evaluated under the risk significance determination process as having very low safety

significance (Green). Additionally, a licensee identified violation is listed in Section 4OA7 of this

report. If you contest this 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 Nuclear Regulatory

Commission, ATTN: Document Control Desk, Washington DC 20555-0001; and the NRC

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

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its

enclosure, and your response 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/

William B. Jones

Project Branch E

Division of Reactor Projects

Entergy Operations, Inc.

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Docket: 50-382

License: NPF-38

Enclosure:

NRC Inspection Report

50-382/03-04

cc w/enclosure:

Executive Vice President and

Chief Operating Officer

Entergy Operations, Inc.

P.O. Box 31995

Jackson, Mississippi 39286-1995

Vice President, Operations Support

Entergy Operations, Inc.

P.O. Box 31995

Jackson, Mississippi 39286-1995

Wise, Carter, Child & Caraway

P.O. Box 651

Jackson, Mississippi 39205

General Manager, Plant Operations

Waterford 3 SES

Entergy Operations, Inc.

17265 River Road

Killona, Louisiana 70066-0751

Manager - Licensing Manager

Waterford 3 SES

Entergy Operations, Inc.

17265 River Road

Killona, Louisiana 70066-0751

Chairman

Louisiana Public Service Commission

P.O. Box 91154

Baton Rouge, Louisiana 70821-9154

Director, Nuclear Safety &

Regulatory Affairs

Waterford 3 SES

Entergy Operations, Inc.

17265 River Road

Killona, Louisiana 70066-0751

Entergy Operations, Inc.

-3-

Michael E. Henry, Administrator

and State Liaison Officer

Department of Environmental Quality

P.O. Box 82135

Baton Rouge, Louisiana 70884-2135

Parish President

St. Charles Parish

P.O. Box 302

Hahnville, Louisiana 70057

Winston & Strawn

1400 L Street, N.W.

Washington, D.C. 20005-3502

Technological Services

Branch Chief

FEMA Region VI

800 North Loop 288

Federal Regional Center

Denton, Texas 76201-3698

Entergy Operations, Inc.

-4-

Electronic distribution by RIV:

Regional Administrator (EWM)

DRP Director (ATH)

DRS Director (DDC)

Senior Resident Inspector (MCH)

Branch Chief, DRP/E (WBJ)

Senior Project Engineer, DRP/E (VGG)

Staff Chief, DRP/TSS (PHH)

RITS Coordinator (NBH)

Brian McDermott (BJM)

WAT Site Secretary (AHY)

W. A. Maier, RSLO (WAM)

Dale Thatcher (DFT)

ADAMS: Yes

No Initials: ______

Publicly Available Nonpublicly Available

Sensitive

Nonsensitive

R:\\_WAT\\2003\\WT2003-04RP-MCH.wpd

RIV:RI:DRP/E

SRI:DRP/E

C:DRS/EMB

C:DRS/PSB

C:DRP/E

GFLarkin

MCHay

CSMarschall

TWPruett

WBJones

T - VGGaddy

/RA/

E - MPShannon /RA/

4/11/03

4/9/03

4/17/03

OFFICIAL RECORD COPY

T=Telephone E=E-mail F=Fax

ENCLOSURE

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket:

50-382

License:

NPF-38

Report:

50-382/03-04

Licensee:

Entergy Operations, Inc.

Facility:

Waterford Steam Electric Station, Unit 3

Location:

Hwy. 18

Killona, Louisiana

Dates:

December 29, 2002, through March 22, 2003

Inspectors:

M. C. Hay, Senior Resident Inspector

G. F. Larkin, Resident Inspector

J. M. Mateychick, Reactor Inspector

P. A. Goldberg, Senior Reactor Inspector

Paul J. Elkmann, Emergency Preparedness Inspector

Approved By:

W. B. Jones, Chief, Project Branch E

Attachment:

Supplemental Information

SUMMARY OF FINDINGS

Waterford Steam Electric Station, Unit 3

NRC Inspection Report 50-382/03-04

IR05000382/2003-04; Entergy Operations, Inc.; on 12/29/2002-03/22/2003; Waterford Steam

Electric Station; Unit 3; Event Followup

The report covered a 12-week period of inspection by resident inspectors, an emergency

preparedness inspector, and a reactor inspector. The inspection identified one Green finding.

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using

IMC 0609, Significance Determination Process (SDP). Findings for which the SDP does not

apply may be Green or be assigned a severity level after NRC management review. The

NRCs program for overseeing the safe operation of commercial nuclear power reactors is

described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

A.

Inspector Identified and Self-Revealing Findings

Cornerstone: Initiating Events

Green. A self-revealing finding was identified for the failure to maintain and

operate main generator seal oil backup differential pressure regulating

Valve SO-308 in accordance with vendor recommendations. This condition

resulted in a turbine trip and subsequent reactor power cutback on

February 14, 2003.

This self-revealing finding is greater than minor because it resulted in a

perturbation in plant stability resulting in a reactor power cutback, similar to

example 4.b in Appendix E of Manual Chapter 0612. The finding is of very low

safety significance because, although it caused a plant transient, it did not

increase the likelihood of a primary or secondary system loss-of-coolant accident

initiator, did not contribute to the loss of mitigation equipment functions, and did

not increase the likelihood of a fire or internal/external flood (Section 4OA3).

B.

Licensee-Identified Violations

A violation of very low safety significance, which was identified by the licensee, was

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

been entered into the licensee's corrective action program. The violation and corrective

action tracking number is listed in Section 4OA7.

Report Details

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

December 29, 2002, through February 14, 2003. Power was reduced to approximately

60 percent power February 14, 2003, following a turbine trip and subsequent reactor power

cutback. The reactor was shutdown February 15, 2003, to support main turbine generator

repairs. On February 19, 2003, reactor power was restored to 100 percent. Reactor power

was maintained at approximately 100 percent throughout the remainder of the inspection

period.

1

REACTOR SAFETY

Initiating Events, Mitigating Systems, Barrier Integrity (R)

1R01

Adverse Weather Protection (71111.01)

a.

Inspection Scope

On January 23, 2003, the inspectors performed a walkdown of components and

systems susceptible to freezing using Procedure OP-002-007, Freeze Protection and

Temperature Maintenance, Revision 10, to verify that the onset of cold weather would

not affect mitigating systems. This inspection included a review of deficiency tags and

condition reports associated with heat tracing and other cold weather protection

measures to determine their impact on the systems. Additionally, the inspectors

discussed adverse weather preparations with various licensee personnel.

b.

Findings

No findings of significance were identified.

1R04

Equipment Alignment (71111.04)

.1

Reactor Auxiliary Building Cable Vault and Switchgear Area Ventilation System

a.

Inspection Scope

The inspectors performed a complete equipment alignment inspection of the reactor

auxiliary building cable vault and switchgear area ventilation system. A review of select

maintenance work orders and corrective action documents was performed to assess the

material condition and performance of the switchgear area ventilation system. System

configuration was assessed using Operating Procedure OP-003-026, "Cable Vault and

Switchgear HVAC," Revision 7. A walkdown of accessible portions of the system was

performed to assess material condition, such as system leaks and housekeeping issues,

that could adversely affect system operability. The inspection also consisted of verifying

that the system was installed, maintained, and tested as described in the Updated Final

Safety Analysis Report and Technical Specifications.

b.

Findings

Introduction: The NRC identified that Switchgear Ventilation System Trains A and B

safety-related outside air intake Dampers SVS-101 and SVS-102, respectively, are

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susceptible to a common mode failure vulnerability associated with a loss of

nonsafety-related instrument air. Pending determination of the findings safety

significance, this finding is identified as Unresolved Item (URI) 50-382/03-04-01.

Description: Switchgear Ventilation System Trains A and B outside air intake

Dampers SVS-101 and SVS-102, respectively, are safety-related, installed in series,

and pneumatically operated. A safety injection actuation signal automatically positions

these dampers to a minimum open position using instrument air. The inspectors noted

that a loss of instrument air, which is a nonsafety-related system, would introduce a

common mode failure for Dampers SVS-101 and SVS-102 preventing these valves from

performing their safety-related function during certain postaccident conditions. In

response to this concern, the licensee took immediate corrective actions and gagged, in

the minimum open position, Damper SVS-102. A review of design documentation by

the inspectors and the licensee identified that the basis for the valves being positioned

in the minimum open position following a safety injection actuation signal was not clearly

documented. The licensee developed, but had yet to implement, a special test to

assess the effects on the control room envelope and those areas surrounding the

control room due to Dampers SVS-101 and SVS-102 failing in the open position.

Analysis: Using the guidance in Appendix B of Inspection Manual Chapter 0612, this

issue potentially will screen more than minor. The barrier integrity objective, to provide

reasonable assurance that the physical design barriers to protect the control room

operators from radionuclide releases caused by accidents or events, was affected.

A Phase 1 screening was performed for the issue utilizing NRC Manual Chapter 0609,

Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Attachment 1. The finding was assessed as potentially affecting the

radiological barrier function for the control room. The significance of this issue is

unresolved pending the results of a special test that will determine the pressure effects

on the control room envelope following failure of Dampers SVS-101 and SVS-102 to

maintain their minimum open safety position following a safety injection actuation signal.

Enforcement: 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in

part, that Measures shall be established to assure that applicable regulatory

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

procedures, and instructions. The failure to maintain design control of the switchgear

ventilation system resulting in the potential common mode failure of Dampers SVS-101

and SVS-102, due to loss of the nonsafety related instrument air system, is being

considered a violation of 10 CFR Part 50, Appendix B, Criterion III. Pending

determination of the findings safety significance, this finding is identified as

URI 50-382/03-04-01. The licensee documented this issue in their corrective action

process as Condition Report CR-WF3-2003-0062.

-3-

.2

High-Pressure Safety Injection System Train A

a.

Inspection Scope

On February 4, 2003, the inspectors performed a partial walkdown of the mechanical

and electrical components of a critical portion of High-Pressure Safety Injection System

Train A while the train was in a standby alignment. This walkdown was completed

during scheduled maintenance that rendered Train B inoperable. The inspectors

verified that the system was installed, maintained, and tested as described in the

Updated Final Safety Analysis Report and Technical Specifications.

b.

Findings

No findings of significance were identified.

.3

Shield Building Ventilation System

a.

Inspection Scope

On January 29, 2003, the inspectors performed a partial walkdown of the mechanical

and electrical components of a critical portion of Shield Building Ventilation System

Train A. This walkdown was completed during scheduled maintenance that rendered

Train B inoperable. The inspectors verified that the system was installed, maintained,

and tested as described in the Updated Final Safety Analysis Report and Technical

Specifications.

b.

Findings

No findings of significance were identified.

.4

Component Cooling Water Train A

a.

Inspection Scope

On March 13, 2003, the inspectors completed a partial equipment alignment inspection

of Component Cooling Water Train A. A review of select maintenance work orders and

corrective action documents was performed to assess the material condition and

performance of Component Cooling Water Train A. System configuration was assessed

using Operating Procedure OP-002-003, "Component Cooling Water," Revision 13. A

walkdown of accessible portions of the system was performed to assess material

condition, such as system leaks and housekeeping issues, that could adversely affect

system operability.

b.

Findings

No findings of significance were identified.

-4-

1R05

Fire Protection (71111.05)

The inspectors conducted six inspections to determine if the licensee had implemented

a fire protection program that adequately controlled combustibles and ignition sources

within the plant, effectively maintained fire detection and suppression capabilities, and

maintained passive fire protection features in good material condition.

The following areas were inspected:

Reactor auxiliary building +21-foot elevation on January 23, 2003

Control room envelop on February 4, 2003

Safety Injection Pump Area B on February 4, 2003

Switchgear Room B on February 4, 2003

Reactor auxiliary building +46-foot elevation on February 20, 2003

Reactor auxiliary building -4-foot and -35-foot elevations on February 28, 2003

b.

Findings

No findings of significance were identified.

1R11

Licensed Operator Requalification (71111.11)

a.

Inspection Scope

On February 3, 2003, the inspectors observed licensed operator simulator training. The

simulator training evaluated the operator's ability to recognize, diagnose, and respond to

a small tube leak in Steam Generator 1, a reactor trip with failure of two control element

assemblies to insert, and the failure of High-Pressure Safety Injection Pump B to start

on a safety injection actuation signal. The inspectors observed and evaluated the

following areas:

Understanding and interpreting annunciator and alarm signals

Diagnosing events and conditions based on signals or readings

Understanding plant systems

Use and adherence of Technical Specifications

Crew communications including command and control

The crew's and evaluator's critiques

b.

Findings

No findings of significance were identified.

-5-

1R12

Maintenance Rule Implementation (71111.12)

.1

Routine Maintenance Rule Review

a.

Inspection Scope

During the inspection period, the inspectors reviewed licensee implementation of the

Maintenance Rule. The inspectors considered the characterization, safety significance,

performance criteria, and appropriateness of goals and corrective actions. The

inspectors assessed the licensees implementation of the Maintenance Rule to the

requirements outlined in 10 CFR 50.65 and Regulatory Guide 1.160, Monitoring the

Effectiveness of Maintenance at Nuclear Power Plants, Revision 2. The inspectors

reviewed the following systems that displayed performance problems:

Emergency Diesel Generating System Train A

Containment Cooling HVAC Trains A and B

b.

Findings

No findings of significance were identified.

.2

Periodic Evaluation Reviews

a.

Inspection Scope

The inspectors reviewed the Waterford 3 report documenting the performance of the

last Maintenance Rule periodic effectiveness assessment. This periodic evaluation

covered the period from November 2000 through April 2002.

The inspectors reviewed the program for monitoring risk-significant functions associated

with structures, systems, and components using reliability and unavailability. The

performance monitoring of nonrisk-significant functions using plant level criteria was

also reviewed.

The inspectors evaluated whether the report contained adequate assessment of the

performance of the Maintenance Rule Program as well as conformance with applicable

programmatic and regulatory requirements. To accomplish this, the inspectors verified

that the licensee appropriately and correctly addressed the following attributes in the

assessment reports:

The program treatment of nonrisk-significant structure, system, and component

functions monitored against plant level performance criteria

Program adjustments made in response to unbalanced reliability and availability

The application of industry operating experience

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Performance review of Category (a)(1) systems

Evaluation of the bases for system category status change (e.g., (a)(1) to (a)(2)

or (a)(2) to (a)(1))

Effectiveness of performance and condition monitoring at component, train,

system, and plant levels

Review and adjustment of definitions of functional failures

The inspector also verified that the issuance of the two most recent assessments met

the regulatory timeliness requirements.

The inspectors reviewed procedures, condition reports, and Category (a)(1) recovery

plans associated with the above activities for the following systems: core protection

calculator, emergency diesel generator sequencer, feedwater, broad range gas

monitors, process radiation monitors, essential chillers (refrigeration), and shutdown

cooling.

b.

Findings

No findings of significance were identified.

.3

Identification and Resolution of Problems

a.

Inspection Scope

The inspectors evaluated the use of the corrective action system within the Maintenance

Rule Program for issues associated with risk-significant systems. The inspectors

examined a sample of corrective action documents associated with systems which were,

or had been, in Maintenance Rule Category (a)(1), including recovery plans for

improving the system performance. The inspectors performed this review to establish

that the corrective action program was entered at the appropriate threshold for the

purpose of:

Implementing the corrective action process when a performance criterion was

exceeded

Correcting performance-related issues or conditions identified during the periodic

evaluation

Correcting generic issues or conditions identified during programmatic

assessments, audits, or surveillances.

The inspectors identified an observation concerning the licensee's implementation of

appropriate corrective actions to maintain the performance of the core protection

calculator system. The core protection calculator was placed in Maintenance Rule

-7-

status Category (a)(1) from July 2000 to February 2001 and again in December 2001

until the time of this inspection due to both functional failures and unavailability.

The inspectors reviewed the licensees goals for monitoring the performance of the core

protection calculator for the Maintenance Rule. The inspectors noted that one

performance criteria for each channel to remain in Category (a)(2) consisted of

functional failures that resulted in a spurious channel trip being < 20 functional failures

per 18-month period per channel. The inspectors found that Channel D is not meeting

this goal. The inspectors review of system performance since July 1999 against this

goal indicated that prior to July 2000 there was another period when Channel D was not

meeting the goal. In addition, the inspectors found that Channels B and C also had

periods of not meeting this goal, sometimes concurrently with Channel D. The licensee

stated that since 1999 there were 13 instances where one channel was in trip and a

second channel was in bypass.

The inspectors reviewed Condition Reports CR-WF3-2000-0839 and -2001-1346 and

found that the licensee's corrective actions were focused on replacing failed electronic

components and improving the ventilation flow through the core protection calculator

cabinets to reduce the operating temperature of the electronic components. The

licensee intends to maintain the current core protection calculators system until

replacement during Refueling Outage 14 in the fall of 2006.

b.

Findings

No findings of significance were identified.

1R13

Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)

a.

Inspection Scope

The inspectors reviewed risk assessments for planned or emergent maintenance

activities to determine if the licensee met the requirements of 10 CFR 50.65(a)(4) for

assessing and managing any increase in risk from these activities. Risk evaluations of

the following five occurrences were reviewed:

On January 23, 2003, Nitrogen Gas Valve NG-709 was declared inoperable and

required emergent repairs.

On February 5, 2003, Emergency Feedwater Pump B was declared inoperable

and required emergent repairs.

On February 6, 2003, troubleshooting activities were performed to isolate a

ground on the control element drive mechanism control system.

On February 9, 2003, Main Steam Admission Valve MS-401B to the

turbine-driven auxiliary feedwater pump was declared inoperable and required

emergent repairs.

-8-

On March 14, 2003, the Plant Protection System Channel B High Log Power Trip

Bypass Module was replaced.

b.

Findings

No findings of significance were identified.

1R14

Personnel Performance During Nonroutine Plant Evolutions (71111.14)

a.

Inspection Scope

For the nonroutine events described below, the inspectors reviewed operator logs, plant

computer data, and strip charts to determine what occurred, how the operators

responded, and whether the response was in accordance with plant procedures:

On February 14, 2003, the inspectors observed the site response to a turbine trip

followed by a reactor power cutback from 100 percent power. Reactor power

was reduced to approximately 60 percent with the steam bypass control system

available to mitigate the transient. On February 15, 2003, the inspectors

observed the operators perform a reactor shutdown following the identification of

insulation degradation that affected the main generator exciter.

b.

Findings

No findings of significance were identified.

1R15

Operability Evaluations (71111.15)

a.

Inspection Scope

The inspectors reviewed the technical adequacy of four operability evaluations to verify

that they were sufficient to justify continued operation of a system or component. The

inspectors considered that, although equipment was potentially degraded, the operability

evaluation provided adequate justification that the equipment could still meet its

Technical Specification, Updated Final Safety Analysis Report, and design-bases

requirements and that the potential risk increase contributed by the degraded equipment

was thoroughly evaluated. The following evaluations were reviewed:

Operability evaluation addressing missing nuts, washers, and a U-bolt affecting

the auxiliary component cooling water system Wet Cooling Tower A (Condition

Report CR-WF3-2003-00089)

Operability evaluation addressing broken reach rod linkage affecting operation of

Containment Spray Valve CS-117B (Condition Report CR-WF3-2003-00309)

-9-

Operability evaluation addressing total component cooling water flow in the

accident alignment exceeding design flow rates (Condition Report CR-WF3-2003-00512)

Operability evaluation addressing degraded seal water flow to Charging Pump B

(Condition Report CR-WF3-2003-00640)

b.

Findings

No findings of significance were identified.

1R16

Operator Workarounds (71111.16)

a.

Inspection Scope

The inspectors performed a review of operator workarounds. This review evaluated the

cumulative affects of current operator workarounds to assess the overall impact

affecting the operators ability to respond in a correct and timely manner to plant

transients and accidents.

b.

Findings

No findings of significance were identified.

1R19

Postmaintenance Testing (71111.19)

a.

Inspection Scope

The inspectors reviewed postmaintenance tests to verify system operability and

functional capabilities. The inspectors considered whether testing met design and

licensing bases, Technical Specifications, and licensee procedural requirements. The

inspectors reviewed the testing results for the following six components:

Essential Chiller A following a low refrigerant pressure trip due to refrigerant

leakage through a damaged dehydrator gasket joint on December 12, 2002

Nitrogen Gas Valve NG-811 following repair work on valve internal parts on

February 20, 2003

Nitrogen Gas Valve NG-709 following valve stroke failure on February 23, 2003

Chilled Water Valve CHW-900 following valve actuator maintenance on

February 25, 2003

Main Steam Valve MS-401B following motor replacement on March 10, 2003

Plant Protection System Channel B High Log Power Trip Bypass Module

following replacement on March 14, 2003

-10-

b.

Findings

No findings of significance were identified.

1R22

Surveillance Testing (71111.22)

a.

Inspection Scope

The inspectors observed or reviewed the following six surveillance tests to ensure the

systems were capable of performing their safety function and to assess their operational

readiness. Specifically, the inspectors considered whether the following surveillance

tests met Technical Specifications, the Updated Final Safety Analysis Report, and

licensee procedural requirements:

Surveillance Procedure OP-903-030, Safety Injection Pump Operability

Verification, Revision 13, was reviewed on January 24, 2003. This surveillance

tested the functional capability of Low-Pressure Safety Injection Pump A.

Surveillance Procedure OP-903-046, Emergency Feedwater Pump Operability

Check, Revision 15, performed on February 5, 2002. This surveillance tested

the functional capability of motor-driven Emergency Feedwater Pump B.

Surveillance Procedure OP-903-107, Plant Protection System Channel

_A_B_C_D Functional Test, Revision 14, was reviewed on February 19, 2003.

This surveillance tested the bypass, pretrip, and trip actuation capability of Plant

Protection System Channel A.

Surveillance Procedure STA-001-001, Containment Air Lock Seal Leakage

Test, Revision 4, was reviewed on February 20, 2003. This surveillance tested

the containment air lock pressure decay rate.

Surveillance Procedure OP-903-102, Safety Channel Nuclear Instrumentation

Functional Test, Revision 10, was reviewed on February 21, 2003. This

surveillance tested the functional capability of the Excore Nuclear Safety

Channels.

Surveillance Procedure OP-903-043, Shield Building Ventilation System

Operability Check, Revision 9, was reviewed on March 10, 2003. This

surveillance tested stroke times for critical valves required to change position

and verified adequate flow rates through the filter media.

b.

Findings

No findings of significance were identified.

-11-

1R23

Temporary Plant Modifications (71111.23)

a.

Inspection Scope

The inspectors reviewed a temporary plant modification of the switchgear ventilation

system to ensure that the modification did not adversely affect system operability or

design requirements specified in the Updated Final Safety Analysis Report and

Technical Specifications. The modification consisted of gagging switchgear ventilation

system Damper SVS-102 in the minimum open position. This modification was installed

to place the damper in its fail safe position after identifying that a loss of

nonsafety-related instrument air would prevent the valve from performing its

safety-related function during certain postaccident conditions. The inspectors reviewed

the following documentation during this inspection activity:

Condition Report CR-WF3-2003-00062

EC-F00-0026, Post Fire Safe Shutdown Analysis

Updated Final Safety Analysis Report, Section 9.4.3, Reactor Auxiliary Building

Ventilation

b.

Findings

No findings of significance were identified.

Emergency Preparedness (EP)

1EP4

Emergency Action Level and Emergency Plan Changes (71114.04)

a.

Inspection Scope

The inspector performed an in-office review of Revision 28 to the Waterford 3

Emergency Plan, submitted January 15, 2003. Revision 28 changed organizational

titles, updated facility and equipment information, clarified the revision process for the

emergency plan and emergency action levels, and made editorial corrections.

Revision 28 also implemented aspects of the removal of the postaccident sampling

system as approved in Technical Specification Amendment 172. The inspector

compared Revision 28 with its previous revision and with the requirements of 10 CFR 50.54(q) to determine if the revision decreased the effectiveness of the emergency plan.

b.

Findings

No findings of significance were identified.

-12-

1EP6

Drill Evaluation (71114.06)

a.

Inspection Scope

The inspectors reviewed the drill scenario and observed activities in the simulated

control room and the emergency operations facility. The drill scenario simulated

equipment failures, a site evacuation, a loss of coolant accident, and the release of

radioactive material offsite. In addition, the inspectors reviewed the drill critiques and

the resolution of identified performance problems. The drill was conducted on

March 13, 2003.

b.

Findings

No findings of significance were identified.

4

OTHER ACTIVITIES (OA)

4OA1 Performance Indicator Verification (71151)

.1

Initiating Events and Barrier Integrity Performance

a.

Inspection Scope

The inspectors reviewed data for initiating events and barrier integrity cornerstone

performance indicators from the fourth quarter of 2001 through the third quarter of 2002

for the following:

Performance indicator data for unplanned power changes per 7,000 critical

hours

Performance indicator data for scrams with loss of normal heat removal

Performance indicator data for safety system unavailability/emergency ac power

b.

Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152)

a.

Inspection Scope

The inspectors reviewed the licensees corrective actions associated with the failure of

Main Steam Admission Valve MS-401B for the turbine-driven emergency feedwater

pump. This valve failed to operate during surveillance testing on March 9, 2003. The

inspectors reviewed Condition Report CR-WF3-2003-00616 to ensure the full extent of

-13-

the issue was identified, appropriate evaluations were performed, and corrective actions

were specified and prioritized. Additionally, the inspectors reviewed maintenance history

on the valve to ensure that maintenance activities were accomplished in accordance

with vendor recommendations and specifications.

b.

Findings

No findings of significance were identified.

4OA3 Event Followup (71153)

a.

Inspection Scope

On February 14, 2003, the plant experienced a main turbine trip and subsequent reactor

power cutback while transferring Electrical Bus 3AB to an alternate power supply. On

February 15, 2003, the reactor was subsequently shut down after identifying insulation

degradation affecting the main generator exciter armature. The inspectors assessed

plant response to the transient conditions resulting from the turbine trip to verify safety

systems performed appropriately. The inspectors reviewed the licensees actions to

identify and correct those degraded conditions that could impact plant restart.

b.

Findings

Introduction: A Green self-revealing finding was identified for the failure to maintain and

operate main generator seal oil backup differential pressure regulating Valve SO-308 in

accordance with vendor recommendations. This condition resulted in a turbine trip and

subsequent reactor power cutback on February 14, 2003.

Description: On February 14, 2003, the licensee transferred Electrical Bus 3AB to an

alternate power supply. The electrical bus transfer resulted in the loss of one of the two

available air side seal oil pumps. During the bus transfer, a turbine trip occurred due to

low generator seal oil differential pressure. The licensees investigation revealed that

seal oil backup differential pressure regulating Valve SO-308 had operated slowly and

was set at an inappropriate pressure that ultimately resulted in the turbine trip. Vendor

recommendations consisted of setting the pressure regulator to a setpoint of 8 psid.

The setpoint for the regulator was found to be set at approximately 3 psid, which was

below the turbine trip setpoint. The licensee also noted that the vendor recommended

monthly cycling of Valve SO-308 to verify its proper operation was never implemented

nor contained in a maintenance instruction.

Analysis: The inspectors determined this finding was more than minor because it

caused a perturbation in plant stability resulting in a reactor power cutback. Although

the finding resulted in a plant transient, the inspectors determined that it did not

contribute to the likelihood of a primary or secondary system loss-of-coolant accident

initiator, did not contribute to the loss of mitigation equipment functions, and did not

increase the likelihood of a fire or internal/external flood. Therefore, the failure to

maintain and operate seal oil backup differential pressure regulating Valve SO-308 in

-14-

accordance with vendor recommendations was of very low safety significance (Green).

The licensee documented this issue in their corrective action process as Condition

Report CR-WF3-2003-0408.

Enforcement: No violation of regulatory requirements occurred. The inspectors

determined that the finding did not represent a noncompliance because it occurred on

nonsafety-related secondary plant equipment.

4OA6 Meetings

Exit Meeting Summary

1.

The reactor inspector presented the inspection results to Mr. Joseph Venable,

Waterford Vice President, and other members of licensee management at the

conclusion of the inspection on January 17, 2003.

2.

The inspector presented the inspection results to Mr. J. Lewis, Emergency Planning

Manager, and other members of licensee management during a telephonic exit interview

conducted on March 18, 2003. The licensee acknowledged the findings presented.

3.

The resident inspectors presented the inspection results to Mr. Joseph Venable,

Waterford Vice President, and other members of licensee management at the

conclusion of the inspection on March 24, 2003. 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, NUREG-1600, for being dispositioned a noncited

violation.

10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that

Measures shall be established to assure that applicable regulatory requirements and

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

instructions. Contrary to this, the licensee identified that Component Cooling Water

Trains A and B total flow rates, in an accident condition, exceeded the maximum

analyzed flow rates. This condition resulted in reducing the efficiency of the dry cooling

towers to remove heat under certain environmental conditions. This was identified in the

licensees corrective action process as Condition Report CR-WF3-2003-0512. This

finding is of very low safety significance because the design control deficiency did not

result in loss-of-system function as described in Generic Letter 91-18.

SUPPLEMENTAL INFORMATION

PARTIAL LIST OF PERSONS CONTACTED

Licensee

S. S. Anders, Superintendent, Plant Security

J. R. Douet, General Manager, Plant Operations

C. Fugate, Assistant Manager, Operations

T. Gaudet, Director, Planning and Scheduling

B. Houston, Superintendent, Radiation Protection

C. Lambert, Director, Engineering

J. Lewis, Emergency Planning Manager

R. Murillo, Acting Manager, Licensing

R. Osborne, Manager, System Engineering

K. Peters, Director, Nuclear Safety Assurance/Emergency Preparedness

J. Laque, Manager, Maintenance

G. Scott, Engineer, Licensing

T. E. Tankersley, Manager, Training

J. Venable, Vice President, Operations

K. T. Walsh, Manager, Operations

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

50-382/0304-01

URI

Design Control of SVS-101 and SVS-102 (Section 1R04.1)

Discussed

Finding

FIN

Failure to Implement Vendor Recommendations

(Section 40A3)

DOCUMENTS REVIEWED

Procedures

Operating Procedure OP-003-026, Cable Vault and Switchgear HVAC, Revision 7

Surveillance Procedure OP-903-046, Emergency Feedwater Pump Operability Check,

Revision 15

Technical Procedure PE-004-024, "ACCW and CCW System Flow Balance," Revision 1

Surveillance Procedure OP-903-102, Safety Channel Nuclear Instrumentation Functional

Test, Revision 10

-2-

Surveillance Procedure OP-903-107, Plant Protection System Channel _A_B_C_D Functional

Test, Revision 14

Surveillance Procedure STA-001-001, Containment Air Lock Seal Leakage Test, Revision 4

Operations Procedure OP-903-063, Chilled Water Operability Verification, Revision 11

Corrective Action Documents

CR 2003-0302,2002-2068, 2002-2097, CR 2003-0512, CR 2002-2073, CR 2003-0656,

and CR 2003-0167

Other

Vendor Technical Manual 457000142, "Zurn Industries Mechanical Draft Cooling Towers,"

Revision 11

Calculation Number MN(Q) 9-52, "UHS Performance Based on Test Data," Revision 1

Calculation Number MN(Q) 9-2, "Component Cooling Water System," Revision 1

Calculation Number EC-M95-008, "Ultimate Heat Sink Design Basis," Revision 1

Information Notice 96-01, "Potential for High Post-Accident Closed-Cycle Cooling Water

Temperatures to Disable Equipment Important to Safety

W3-DBD-037, "Essential Chilled Water System Design Bases Document," Revision 1

Technical Procedure PE_004-026, "HVC-101 and HVC-102 Leak Test," Revision 6

Calculation Number NOSG-LPLK-90-01, "Control Room Habitability," Revision 0

W3-DBD-038, "Safety Related HVAC - Control Room Design Bases Document," Revision 1

Calculation Number EC-S96-011, "LOCA Offsite and Control Room Radiological Dose

Consequences," Revision 1

Calculation Number EC-S97-025, "Control Room Habitability Following Accidental Chlorine

Release," Revision 1

Maintenance Action Items

429807, 433429, 438626

-3-

Work Order Package

50231536, 00023206, 50231499, 50088469, 50010906, 00019905, 00023866, and 00020353

Condition Reports

CR-WF3-1998-00250

CR-WF3-2000-00845

CR-WF3-2001-00775

CR-WF3-2001-00863

CR-WF3-2001-00917

CR-WF3-2001-01344

CR-WF3-2001-01347

CR-WF3-2002-00358

CR-WF3-2003-00051

CR-WF3-2003-00053

CR-WF3-2003-00069

Engineering Reports

NUMBER

DESCRIPTION

REVISION

Maintenance Rule Periodic (a)(1) Assessment

Cycle 11

Maintenance Rule Periodic (a)(1) Assessment

Cycle 10

Licensee Event Reports

NUMBER

DESCRIPTION

REVISION

01-001

Violation of TS 3.3.1 because a TS channel check

0

was not performed as required by TS 4.3.1.1

01-003

Reactor Protection System Trip caused by Turbine

0

Governor Valve Oscillation

01-004

Failure to enter TS action statement due to inadequate

0

surveillance test procedure

Procedures

NUMBER

DESCRIPTION

REVISION

DC-121

Maintenance Rule

0

-4-

Miscellaneous Documents

NUMBER

DESCRIPTION

REVISION

Entergy South Maintenance Rule Desktop

1

Core Protection Calculators (CPCs) Top Ten

1/9/03

Equipment Issues Plan

Expert Panel Meeting Minutes

8/13/98

Expert Panel Meeting Minutes

11/07/01

LIST OF ACRONYMS USED

CFR

Code of Federal Regulations

FIN

finding

NRC

U. S. Nuclear Regulatory Commission

URI

unresolved item

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