ML031070497
| 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) | |
See also: IR 05000382/2003004
Text
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. -2-
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 T - VGGaddy /RA/ E - MPShannon /RA/
4/11/03 4/11/03 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 A, 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
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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.
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- 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)
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- 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
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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.
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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.
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.
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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
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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
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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
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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
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Work Order Package
50231536, 00023206, 50231499, 50088469, 50010906, 00019905, 00023866, and 00020353
Condition Reports
CR-WF3-1996-0686 CR-WF3-1996-00870 CR-WF3-1998-00250
CR-WF3-1998-0591 CR-WF3-1999-00701 CR-WF3-2000-0698
CR-WF3-2000-00839 CR-WF3-2000-00845 CR-WF3-2000-00855
CR-WF3-2001-00775 CR-WF3-2001-00858 CR-WF3-2001-00863
CR-WF3-2001-00900 CR-WF3-2001-00917 CR-WF3-2001-01112
CR-WF3-2001-01344 CR-WF3-2001-01346 CR-WF3-2001-01347
CR-WF3-2002-00900 CR-WF3-2002-00358 CR-WF3-2002-0563
CR-WF3-2002-01596 CR-WF3-2003-00051 CR-WF3-2003-00052
CR-WF3-2003-00053 CR-WF3-2003-00056 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
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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