ML070450251
| 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, 2007Kevin WalshVice President Operations
Waterford 3
Entergy Operations, Inc.
17265 River Road
Killona, LA 70066-0751SUBJECT:WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATEDINSPECTION REPORT 05000382/2006005 Dear Mr. Walsh:
On December 31, 2006, the NRC completed an inspection at your Waterford Steam ElectricStation, 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 andcompliance with the Commission's 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 safetysignificance (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 NRC's "Rules of Practice," a copy of this letter, itsenclosures, 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 NRC's 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, ChiefProject Branch E
Division of Reactor ProjectsDocket: 50-382License: NPF-38Enclosure: 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 39286-1995Vice President, Operations SupportEntergy Operations, Inc.
P.O. Box 31995
Jackson, MS 39286-1995General Manager, Plant OperationsWaterford 3 SES
Entergy Operations, Inc.
17265 River Road
Killona, LA 70057-3093Manager, LicensingEntergy Operations, Inc.
17265 River Road
Killona, LA 70057-3093ChairmanLouisiana Public Service Commission
P.O. Box 91154
Baton Rouge, LA 70825-1697
Entergy Operations, Inc.-3-Director, Nuclear Safety AssuranceEntergy Operations, Inc.
17265 River Road
Killona, LA 70057-3093Richard Penrod, Senior Environmental Scientist, State Liaison Officer
Office of Environmental Services
Northwestern State University
Russsell Hall, Room 201
Natchitoches, LA 71497Parish President
Council
St. Charles Parish
P.O. Box 302
Hahnville, LA 70057ChairpersonDenton 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-SensitiveR:\_REACTORS\_WAT\2006\WT2006-05RP-GFL.wpdRI:DRP/ESRI:DRP/EC:DRS/EB1C:DRS/OBC:DRS/EB2DHOverlandGFLarkinWBJonesATGodyLJSmithT-JACT-JAC/RA//RA//RA/2/5/072/5/072/5/072/8/072/8/07C:PSBC:DRP/EMPShannonJAClark/RA//RA/2/8/072/13/07OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket:50-382License 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 InspectorD. 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
BranchApproved By:Jeff A. Clark, Chief, Project Branch EATTACHMENTS: Supplemental Information
EnclosureSUMMARY OF FINDINGSIR05000382/2006-005; 10/08/2006-12/31/2006; Waterford Steam Electric Station, Unit 3;Postmaintenance Testing; Refueling and Other Outage Activities; Emergency Plan Biennial
Program InspectionThe report covered a 3-month period of inspection by resident inspectors, a project engineer, asenior 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 NRC's 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 TechnicalSpecification 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 pressurizer's 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 becomea 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-2-Cornerstone: Mitigating Systems*Green. A self-revealing violation of very low safety significance of 10 CFRPart 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 licensee's corrective action
program.This finding is greater than minor because it affects the Mitigating Systemscornerstone 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) forfailure 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 licensee's emergency plan. The licensee's failure to conduct the
drill is a performance deficiency because the licensee identified the drill's
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 providingappropriate 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).
Enclosure-3-B.Licensee-Identified ViolationsViolations of very low safety significance, which were identified by the licensee havebeen reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensee's corrective action program. These violations and
corrective actions are listed in Section 4OA7 of this report.
Enclosure-4-REPORT DETAILSSummary of Plant Status: The plant was operated at approximately 100 percent power fromOctober 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 SAFETYCornerstones: Initiating Events, Mitigating Systems, Barrier Integrity1R01Adverse Weather Protection (71111.01) .1Readiness For Seasonal Susceptibilities a.Inspection ScopeThe inspectors completed a review of the licensee's readiness of seasonalsusceptibilities involving low seasonal temperatures and high winds. The inspectors:
(1) reviewed plant procedures, the Updated Final Safety Analysis Report, and TechnicalSpecifications 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 EmergencyFeedwater 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.FindingsNo findings of significance were identified.
Enclosure-5-1R04Equipment Alignment (71111.04) .1Partial Walkdown a.Inspection ScopeThe inspectors: (1) walked down portions of the two below listed risk important systemsand 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 licensee's 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 inspectorscompleted two samples. b.FindingsNo findings of significance were identified.1R05Fire Protection (71111.05) .1Quarterly Inspection a.Inspection ScopeThe inspectors walked down the six below listed plant areas to assess the materialcondition 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
Enclosure-6-*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 ADocuments reviewed by the inspectors are listed in the attachment. The inspectorscompleted six samples. b.FindingsNo findings of significance were identified.1R06Flood Protection Measures (71111.06) .2Semi-annual Internal Flooding a.Inspection ScopeThe inspectors: (1) reviewed the Updated Final Safety Analysis Report, the floodinganalysis, 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.FindingsNo findings of significance were identified.
Enclosure-7-1R11Licensed Operator Requalification Program (71111.11) .1Biennial Inspection a.Inspection ScopeThe inspector reviewed the annual operating examination test results for 2006. Sincethis 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.
.2Quarterly Inspection a.Inspection ScopeOn November 6-7, 2006, the inspectors observed training of senior reactor operatorsand reactor operators to identify deficiencies and discrepancies in the training, to assess
operator performance, and to assess the evaluator's 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 ofEmergency Conditions," Revision 21The inspectors completed one sample. b.FindingsNo findings of significance were identified.
Enclosure-8-1R12Maintenance Effectiveness (71111.12) a.Inspection ScopeThe inspectors reviewed the two below listed Maintenance Rule scoped systems thathave 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 inspectorscompleted two samples. b.FindingsNo findings of significance were identified.1R13Maintenance Risk Assessments and Emergent Work Control (71111.13) .1Risk Assessment and Management of Risk a.Inspection ScopeThe 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 duringreactor coolant system midloop conditions*December 7, 2006: Planned maintenance activities to restore emergency dieselgenerator Train B following the integrated diesel test*December 18, 2006: Planned maintenance and operational activities duringreactor coolant system midloop conditions
Enclosure-9-Documents reviewed by the inspectors are listed in the attachment. The inspectorscompleted four samples. b.FindingsNo findings of significance were identified.1R15Operability Evaluations (71111.15) a.Inspection ScopeThe 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 capacityfollowing a surveillance test as described in Condition Report 2006-3125*November 11, 2006: Operability evaluation addressing containment fan coolerTrain 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 tubesheet plugging as described in Condition Report 2006-4510Documents reviewed by the inspectors are listed in the attachment. The inspectors
completed three samples. b.FindingsNo findings of significance were identified.1R19Postmaintenance Testing (71111.19) a.Inspection ScopeThe inspectors selected the six below listed postmaintenance test activities of risksignificant 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
Enclosure-10-reviewed test data to verify that acceptance criteria were met, plant impacts wereevaluated, 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 chargingPump B*October 12, 2006: Planned maintenance for reactor trip circuit breakerNumber 7*November 1, 2006: Planned maintenance for start-up transformer Train B6.9 kV circuit breaker*November 26, 2006: Emergent maintenance on reactor coolant Loop 1shutdown 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 shutdowncooling suction outside containment isolation Valve SI-407BDocuments reviewed by the inspectors are listed in the attachment. The inspectorscompleted six samples. b.FindingsIntroduction. 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 licensee's corrective action program.Description. On November 26, 2006, while in Mode 4, station operators performedOperating 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
Enclosure-11-attempt, Valve SI-405B fully opened, but not within the maximum inservice testing limitof 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 fullyopen 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 strokeopen 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 establishcorrective 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," requiresin 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
Enclosure-12-quality, the measures shall assure that the cause of the condition is determined andcorrective 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).1R20Refueling and Other Outage Activities (71111.20)The inspectors reviewed the following risk significant refueling items or outage activitiesto 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.FindingsIntroduction. A self-revealing violation of very low safety significance of TechnicalSpecification 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 pressurizer's
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 pressurizercubicle, 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
Enclosure-13-in the event of a safe shutdown earthquake. A steel bolted flange is attached to thebottom 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 ChangeProcedure 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 thatthe 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 toestablish 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 proceduresshall 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.
Enclosure-14-Contrary to these requirements, the licensee failed to ensure that Field ChangeRequest 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
licensee's corrective action program as Condition Report CR-WF3-2006-4274.
(NCV 05000382/2006005-02, Excess Torque Resulting in Pressurizer Skirt Bolt
Failures).
1R22Surveillance Testing (71111.22) a.Inspection ScopeThe 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 ofAir 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 DieselGenerator," 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 SafetyValve 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 FlowBalance 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 inspectorscompleted four samples.
Enclosure-15- b.FindingsNo findings of significance were identified.1R23Temporary Plant Modifications (71111.23) a.Inspection ScopeThe 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 toAugment 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 PartsMonitoring 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 inspectorscompleted two samples. b.FindingsNo findings of significance were identified.
Cornerstone: Emergency Preparedness (EP)1EP2Alert Notification System Testing (71114.02) a.Inspection ScopeThe inspector discussed with licensee staff the status of offsite siren and tone alert radiosystems to determine the adequacy of licensee methods for testing the alert and
notification system in accordance with 10 CFR Part 50, Appendix E. The licensee's alert
and notification system testing program was compared with criteria in NUREG-0654,
Enclosure-16-"Criteria for Preparation and Evaluation of Radiological Emergency Response Plans andPreparedness 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 licensee's 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
- Emergency Preparedness Desk Guide 16, "Siren System Administrative Data,"Revision 11The inspector completed one sample during the inspection. b.FindingsNo findings of significance were identified.1EP3Emergency Response Organization Augmentation Testing (71114.03) a.Inspection ScopeThe inspector discussed with licensee staff the status of primary and backup systemsfor 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
licensee's 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.FindingsNo findings of significance were identified.1EP4Emergency Action Level and Emergency Plan Changes (71114.04) a.Inspection ScopeThe inspector performed an in-office review of Revision 33 to the Waterford SteamElectric Station Emergency Plan, and Revision 21 to Emergency Plan Implementing
Procedure EP-001-001, "Recognition and Classification of Emergency Conditions."
Enclosure-17-These revisions implemented an NEI 99-01, "Methodology for Development ofEmergency 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 Preparationand 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.FindingsNo findings of significance were identified.1EP5Correction of Emergency Preparedness Weaknesses and Deficiencies (71114.05) a.Inspection ScopeThe inspector reviewed the licensee's corrective action program requirements inProcedures 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 licensee's 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 licensee's 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 licensee's audit program requirements inProcedure 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.FindingsIntroduction: A Green noncited violation was identified for failure to conduct a requiredoffsite medical drill during calendar year 2005, as required by 10 CFR 50.54(q).
Enclosure-18-Description: The NRC identified a performance deficiency related to conduct of thelicensee's 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 licensee's 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 licensee's
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 17
th should be postponed. . . .," and ". . . we will be contacting yourorganization 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 licensee's 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 WestJefferson 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 medicaldrill 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
Enclosure-19-licensees's ability to foresee, correct, and prevent. The finding had a credible impact onthe 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, inpart, "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, "...periodicdrills 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 thatemployees 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 thelocal 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 licensee's 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 SAFETYOccupational Radiation Safety (OS)2OS1Access Control to Radiologically Significant Areas (71121.01) a.Inspection ScopeThis area was inspected to assess the licensee's performance in implementing physicaland 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 licensee's
procedures required by Technical Specifications as criteria for determining compliance.
Enclosure-20-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 reportedby 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 samplerlocations*Conformity of electronic personal dosimeter alarm set points with surveyindications 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 airborneradioactivity areas*Adequacy of the licensee's internal dose assessment for any actual internalexposure greater than 50 millirem Committed Effective Dose Equivalent*Physical and programmatic controls for highly activated or contaminatedmaterials (nonfuel) stored within spent fuel and other storage pools *Self-assessments, audits, and special reports related to the access controlprogram since the last inspection*Corrective action documents related to access controls
- Licensee actions in cases of repetitive deficiencies or significant individualdeficiencies *Radiation work permit briefings and worker instructions
- Adequacy of radiological controls such as, required surveys, radiation protectionjob coverage, and contamination controls during job performance *Dosimetry placement in high radiation work areas with significant dose rategradients*Changes in licensee procedural controls of high dose rate - high radiation areasand very high radiation areas*Controls for special areas that have the potential to become very high radiationareas during certain plant operations*Posting and locking of entrances to all accessible high dose rate - high radiationareas and very high radiation areas
Enclosure-21-*Radiation worker and radiation protection technician performance with respect toradiation protection work requirements The inspector completed 17 of the required 21 samples. b.FindingsNo findings of significance were identified.2OS2ALARA Planning and Controls (71121.02) a.Inspection ScopeThe inspector assessed licensee performance with respect to maintaining individual andcollective radiation exposures as low as is reasonably achievable (ALARA). The
inspector used the requirements in 10 CFR Part 20 and the licensee's 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 associatedwork activity exposure estimates, which were likely to result in the highest
personnel collective exposures*Site specific ALARA procedures
- Interfaces between operations, radiation protection, maintenance, maintenanceplanning, scheduling, and engineering groups*Integration of ALARA requirements into work procedure and radiation work permitdocuments*Exposure tracking system
- Workers use of the low dose waiting areas
- First-line job supervisors' contribution to ensuring work activities are conducted ina dose efficient manner*Source-term control strategy or justifications for not pursuing such exposurereduction initiatives*Specific sources identified by the licensee for exposure reduction actions andpriorities established for these actions, and results since the last refueling cycle *Radiation worker and radiation protection technician performance during workactivities 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.
Enclosure-22- b.FindingsNo findings of significance were identified.4.OTHER ACTIVITIES (OA)
4OA1Performance Indicator Verification (71151) .1Mitigating Systems and Barrier Integrity a.Inspection ScopeThe inspectors sampled licensee submittals for the three performance indicators listedbelow 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 licensee's 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 LeakageDocuments reviewed by the inspectors are listed in the attachment. The inspectorscompleted three samples.
b.FindingsNo findings of significance were identified. .2(Closed) Temporary Instruction 2515/169: Mitigating Systems Performance IndexVerification a.Inspection ScopeDuring this inspection period, the inspectors completed a review of the licensee'simplementation of the mitigating systems performance index in accordance with the
guidance provided in Temporary Instruction 2515/169. The review examined the
licensee's 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 indexunavailability 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
Enclosure-23-2005 to 2006 was also reviewed to ensure the licensee properly accounted for the actualunavailability 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.FindingsNo findings of significance were identified. This completes the inspection requirementsfor this Temporary Instruction. .3Emergency Preparedness a.Inspection ScopeThe inspector reviewed licensee evaluations for the three Emergency Preparednesscornerstone 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 licensee's evaluations for each performance
indicator reported during the assessment period. The inspector reviewed 100 percent of drill and exercise scenarios and licensed operatorsimulator 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.FindingsNo findings of significance were identified. .4Occupational Radiation Safety and Public Radiation Safety a.Inspection ScopeThe inspector sampled licensee submittals for the performance indicators listed below forthe period from April 2006 through October 2006. To verify the accuracy of the
performance indicator data reported during that period, performance indicator definitions
Enclosure-24-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 identifiedoccurrences 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 identifiedoccurrences 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.FindingsNo findings of significance were identified.4OA2Identification and Resolution of Problems (71152)
.1Routine Review of Identification and Resolution of Problems a.Inspection ScopeThe inspectors performed a daily screening of items entered into the licensee's correctiveaction 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.FindingsNo findings of significance were identified.
Enclosure-25- .2Selected Issue Follow-up Inspection a.Inspection ScopeIn addition to the routine review, the inspectors selected the one below listed issue for amore in-depth review. The inspectors considered the following during the review of the
licensee's 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 inspectorscompleted one sample. b.FindingsNo findings of significance were identified. .3Semiannual Trend Review a.Inspection ScopeThe inspectors completed a semiannual trend review of repetitive or closely relatedissues 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 licensee's 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 licensee's quarterly trend reports. Corrective
actions associated with a sample of the issues identified in the licensee's trend report
were reviewed for adequacy. Documents reviewed by the inspectors are listed in the attachment. The inspectorscompleted one sample. b.FindingsNo findings of significance were identified.
Enclosure-26- .4Annual Sample Review a.Inspection ScopeThe 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 licensee's problemidentification 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 ObservationsNo findings of significance were identified.4OA3Event Follow-up (71153) .1(Closed) LER 05000382/2005-005-00: Manual Reactor Trip Upon Loss of All CirculatingWater Pumps and Lowering Condenser VacuumOn November 11, 2005, the licensee manually tripped the reactor due to lowering maincondenser 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 andRegulatory Guide 1.45On 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.
Enclosure-27-4OA5Other Activities .1Institute of Nuclear Power Operations (INPO) Audit and Evaluation ReviewThe inspectors completed a review of the INPO audit and evaluation report for EntergyOperation's 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 ScopeThe inspectors reviewed Waterford 3's implementation of plant modifications andprocedure changes committed to in their response to Generic Letter 2004-02. The inspectors observed installation of the containment recirculation sump strainers andrelocation 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 containmentrecirculation 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.FindingsNo findings of significance were identified. This completes the inspection requirementsfor this Temporary Instruction.4OA6Meetings, Including ExitExit Meeting Summary .1On September 25, 2006, the operations inspector discussed the inspection results of thelicensed 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. .2On December 1, 2006, the health physics inspector presented the OccupationalRadiation 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.
Enclosure-28- .3On December 8, 2006, the inspectors presented the results of TemporaryInstruction 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. .4On 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. .5On January 16, 2007, the resident inspectors presented the inspection results toMr. 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. 4OA7Licensee-Identified ViolationsThe following violation of very low safety significance (Green) was identified by thelicensee 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 in1994, 2001, and 2003, and corresponding changes to emergency action levels
were not made. This was identified in the licensee's 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
AttachmentA-1SUPPLEMENTAL INFORMATIONKEY POINTS OF CONTACTLicensee PersonnelS. Anders, Superintendent, Plant SecurityJ. 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 EngineerITEMS OPENED, CLOSED, AND DISCUSSEDOpened05000382/2006-005-01NCVRecurring Failure of Valve SI-405B to Open(Section 1R19)05000382/2006-005-02NCVExcess Torque Resulting in Pressurizer Skirt BoltFailures (Section 1R20) 05000382/2006-005-03NCVFailure to Conduct a Required Offsite medical Drill in2005 (Section 1EP5)
Closed05000382/2006-005-01NCVRecurring Failure of Valve SI-405B to Open(Section 1R19)05000382/2006-005-02NCVExcess Torque Resulting in Pressurizer Skirt BoltFailures (Section 1R20) 05000382/2005-002-01LERRCS Leakage Detection Instrumentation and RegulatoryGuide 1.4505000382/2005-005-00LERManual Reactor Trip Upon Loss of All Circulating WaterPumps and Lowering Condenser Vacuum05000382/2006-005-03NCVFailure to Conduct a Required Offsite medical Drill in2005 (Section 1EP5)
AttachmentA-2LIST OF DOCUMENTS REVIEWEDSection 1R04: Equipment Alignment (71111.04)ProceduresNumberTitleRevisionOP-002-005Chemical and Volume ControlRevision 21Miscellaneous DocumentsUpdated Final Safety Analysis ReportFlow Diagram - Chemical and Volume Control System, G168, Sheet 2, Rev. 48Section 1R05: Fire Protection (71111.05)ProcedureNUMBERTITLEREVISIONAdministrative ProcedureUNT-005-013Fire Protection Program9Operating Procedure 009-004Fire Protection11-8Maintenance Procedure MM-007-010Fire Extinguisher Inspection andExtinguisher Replacement
13AdministrativeProcedure UNT-005-013Fire Protection Program9Fire Protection Procedure FP-001-015Fire Protection SystemImpairments
17Fire Protection Procedure FP-001-017Transient Combustibles19Training Manual ProcedureNTP-202Fire Protection Training11-4Section 1R12: Maintenance Rule (71111.12)ProceduresNumberTitleRevisionDC-121Maintenance Rule1
NUMARC 93-01Industry Guideline for Monitoring the Effectivenessof Maintenance at Nuclear Power Plants
3
AttachmentA-3Condition ReportsCR-WF3-2005-3084CR-WF3-2005-3270
CR-WF3-2006-0346CR-WF3-2006-0395CR-WF3-2006-0397
CR-WF3-2006-1297CR-WF3-2006-1926CR-WF3-2006-2247
CR-WF3-2006-3285Miscellaneous DocumentsEngineering Report W-SE-2005-001Waterford 3 Maintenance Rule Periodic (a)(3)
Assessment
0Section 1R15: Operability EvaluationsProcedures:NUMBERTITLEREVISIONEN-OP-104Operability Evaluation1
OP-903-097Pressurizer Heater Capacity Verification8
OP-035-000Notification Matrix6Miscellaneous DocumentsNUMBERTITLE/SUBJECTREVISIONWaterford 3 LER2003-001-000Loose Breaker Fuse Rendered One Bank of PressurizerProportional Heaters Inoperable
0Condition ReportsCR-WF3-2006-3125CR-WF3-2006-3128
CR-WF3-2000-0228CR-WF3-2006-4501CR-WF3-1997-2491
CR-WF3-2006-2418CR-WF3-1997-2226CR-WF3-2002-1757
CR-WF3-2003-0827Section 1R19: Postmaintenance Testing (71111.19)
AttachmentA-4ProceduresNUMBERTITLEREVISIONMI-003-115Startup and Control Drawer Calibration Channel1 or 2 3MI-012-012Removal and Installation of Excore Detectors3OP-903-101Startup Channel Functional Test Channel 1 and
2 6OP-903-033Cold Shutdown IST Valve Tests20OP-903-003Charging Pump operability CheckRev. 11.Change 1ENS-MA-114Post Maintenance Testing5Miscellaneous DocumentsNUMBERTITLE/SUBJECTREVISIONCEP-IST-1IST Bases Document3Condition ReportsCR-WF3-2006-3610CR-WF3-2005-1362CR-WF3-2000-1347Work Orders44779-01, 65817-01, 73483, ProceduresNumberTitleRevisionOP-903-003Charging Pump Operability CheckRevision 11,Change 1OP-903-121Safety Systems Quarterly IST Valve Tests7OP-903-127Reactor Trip Circuit Breaker PostMaintenance Test
3OP-903-013Monthly Channel Checks14OP-903-011High Pressure Safety Injection PumpPreservice Operability Check
9
AttachmentA-5CEP-IST-1IST Bases Document3Work Orders50947-1, 51041014-01, 63598-01, Section 1R23: Temporary Plant Modifications (71111.23)ProcedureNUMBERTITLEREVISIONSEN-LI-113Licensing Basis Document ChangeProcess 1Miscellaneous DocumentsNUMBERTITLE/SUBJECTREVISIONER-W3-2006-0264-
000Temporary Air Compressors to Augment Station Air0NRC InformationNotice 2004-17Loose Part Detection and Computerized Eddy CurrentData Analysis in Steam Generators
000Install Loose Parts Monitoring Sensors on the SteamGenerators
0Condition ReportsCR-WF3-2006-4524Work Order
84381Section 1EP3: Emergency Response Organization Augmentation Testing (71114.03)EP-002-015, "Emergency Responder Activation," Revision 8EP-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
AttachmentA-6September 12, 2005Drill 2006-07, "Backup Emergency Response Organization Pager Code Drill," conductedNovember 14, 2006Evaluation 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 NuclearEnergy Institute 99-01, Entergy Operations Inc., Waterford Steam Electric Station, Unit 3," dated
June 20, 2005Section 1EP5: Correction of Emergency Preparedness Weaknesses andDeficiencies (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 RespiratoryEquipment and Reviews of CR-ECH-2004-00096 and CR-ECH-2004-0389"Evaluation Reports for:Drill 2005-01, conducted February 17, 2005Drill 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, "1
st and 2 nd Quarter 2005 Roll-Up Assessment: Emergency PlanningDepartment"LO-WLO-2005-0082, "3
rd Quarter 2005 Roll-Up Assessment: Emergency Planning Department"
AttachmentA-7LO-WLO-2005-0106, "4
th Quarter 2005 Roll-Up Assessment: Emergency Planning Department"LO-WLO-2006-0044, "1
st Quarter 2006 Roll-Up Assessment: Emergency Planning Department"LO-WLO-2006-0069, "2
nd Quarter 2006 Roll-Up Assessment: Emergency Planning Department"LO-WLO-2006-0103, "3
rd 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, 2005Assessment Report LO-WLO-2006-00, "Emergency Response Organization Staffing," September25, 2006
Assessment Report LO-WLO-2006-041, "Emergency Planning Performance Indicator
Assessment," April 21, 2006Assessment Report LO-WLO-2006-000, "Emergency Planning Performance IndicatorAssessment," October 29, 2006EN-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, -4418Section 2OS1: Access Controls to Radiologically Significant Areas (71121.01) Corrective Action Documents2006-03645, 2006-03661, 2006-03721,2006-03736, 2006-03739, 2006-03741, 2006-03812, ProceduresENS-RP-102,Radiological Control, Revision 0ENS-RP-105, Radiation Work Permits, Revision 7
ENS-RP-106, Radiological Survey Documentation, Revision 0Audits and AssessmentsLO-WLO-2006-00067-01, Access Control to Radiologically Significant AreasRadiation Work PermitRWP-2006-0005, Tours and InspectionsRWP-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
AttachmentA-8RWP-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)ProceduresCE-002-006, Maintaining Reactor Coolant Chemistry, Revision 13EN-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 7Section 4OA1: Performance Indicator Verification (71151)ProceduresEN-LI-114, Performance Indicator Process, Revision 1Miscellaneous DocumentsQA/Oversight Observations, November 28, 2008RF-14 Actual RCS Cleanup, November 28, 20084OA2Identification and Resolution of Problems (71152)
ProcedureNUMBERTITLEREVISIONSOP-903-094ESTAS Subgroup Relay Test - Operating10
EN-LI-113Licensing Basis Document ChangeProcess 1EN-LI-102Corrective Action Process7OP-002-004Chilled Water System12
OP-002-003Component Cooling Water System12 and 13
EP-001-001Recognition and Classification ofEmergency Conditions20-2 and 21EP-002-010Notifications and Communications30EP-002-052Protective Action Guidelines19
AttachmentA-9Miscellaneous DocumentsNUMBERTITLE/SUBJECTREVISIONER-W3-00-0541-00Evaluate the Essential Chilled Water LeavingTemperature Setpoint
0Quarterly TrendReport 2 nd Quarter 20060Desk Guide 17Drill Control Team Documentation2Training Evaluation Action Request 2006-1230
Waterford Steam Electric Station Emergency Plan33
Operations Department Performance Indicator2006Condition ReportsCR-WF3-1993-0265CR-WF3-1993-0289
CR-WF3-1996-0084CR-WF3-1996-1852CR-WF3-1997-0028
CR-WF3-2000-1553CR-WF3-2002-1876CR-WF3-2006-3402
CR-WF3-2006-34024OA5 Other: NRC Temporary Instructions 2515/166CalculationNUMBERTITLEREVISIONGENE-0000-0054-9349SIS Sump Strainer Stress Report0Engineering RequestsNUMBERTITLEREVISIONER-W3-2003-0394-001Safety Injection SumpModifications
0DrawingsNUMBERTITLEREVISION06-594, Sht. #1Waterford 3 Containment &Strainers A06-595, Sht. #2Waterford 3 Containment &Strainers A06-596, Sht. #3Waterford 3 Containment &A
AttachmentA-10Strainers06-597, Sht. #4Waterford 3 Containment &Strainers A06-598, Sht. #5Waterford 3 Containment &Strainers A06-486, Sht. #0LT-SI-7145AS & LT-SI-7145BSLocal Mounts
4B430, Sht. X-23J-45Instrument Installation Details1B430, Sht. X-23D-8AInstrument Installation Details2B430, Sht. X-23J-25AInstrument Installation Details5
B430, Sht. X-23J-44Instrument Installation Details2
B430, Sht. X-23J-28Instrument Installation Details8Condition ReportsCR-WF3-2006-03273MiscellaneousNUMBERTITLEREVISION/DATELPL-EQMI-08.01Environmental QualificationMaintenance Input for Rosemount
Model 1153 Series A, B &D, 1154 &
1154 Series H Transmitters and
1159 Remote Seals
9SQ-IC-03Rosemount Pressure Transmitters13LPL-EQA-08.01BEnvironmental QualificationAssessment for the Rosemount
1154 Transmitters Used in the
Waterford SES Unit No. 3
3LPL-EQA-08.01FEnvironmental QualificationAssessment for the Rosemount
1159 Remote Seal System Used in
the Waterford SES Unit No. 3
0NOECP-333, Att. 7.1Construction Material TestingDocument, Grout/Mortar Data, Lab
No. W2576, WO No. 76399
2