ML093230675
| ML093230675 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/18/2009 |
| From: | Clark J A NRC/RGN-IV/DRP/RPB-E |
| To: | Kowalewski J Entergy Operations |
| References | |
| IR-09-004 | |
| Download: ML093230675 (35) | |
See also: IR 05000382/2009004
Text
UNITED STATESNUCLEAR REGULATORY COMMISSIONREGION IV612 EAST LAMAR BLVD, SUITE 400ARLINGTON, TEXAS 76011-4125
November 18, 2009
Joseph Kowalewski, Vice President, Operations
Entergy Operations, Inc.
Waterford Steam Electric Station, Unit 3
17265 River Road Killona, LA 70057-3093
Subject: WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000382/2009-004
Dear Mr. Kowalewski: On October 7, 2009, the U.S. Nuclear Regulatory Commission completed an inspection at your Waterford Steam Electric Station, Unit 3. The enclosed integrated inspection report documents the inspection findings, which were discussed on October 1, 2009 , with you and other members of your staff. The inspections examined activities conducted under your license as they relate to safety and compliance with the 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 one NRC identified finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requirements. However, because of the
very low safety significance and because it was entered into your corrective action program, the
NRC is treating this finding as a noncited violatio
n, consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest the violation or the significance of the noncited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,
Washington, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear
Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas,
76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Waterford Steam Electric Station, Unit 3 facility. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with
the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC
Resident Inspector at Waterford Steam Electric Station, Unit 3. The information you provide will be considered in accordance with Inspection Manual chapter 0305.
Entergy Operations, Inc. - 2 - In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its enclosure, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of 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/
Jeffrey A. Clark, P.E.
Chief, Project Branch E Division of Reactor Projects
Docket: 50-382
License: NPF-38
Enclosure:
NRC Inspection Report 05000382/2009004
w/Attachment: Supplemental Information
cc w/Enclosure: Senior Vice President
Entergy Nuclear Operations
P.O. Box 31995 Jackson, MS 39286-1995 Senior Vice President and Chief Operating Officer Entergy Operations, Inc. P.O. Box 31995 Jackson, MS 39286-1995 Vice President, Operations Support
Entergy Services, Inc.
P.O. Box 31995 Jackson, MS 39286-1995 Senior Manager, Nuclear Safety
and Licensing
Entergy Services, Inc. P.O. Box 31995 Jackson, MS 39286-1995
Site Vice President
Waterford Steam Electric Station, Unit 3
Entergy Operations, Inc.
17265 River Road Killona, LA 70057-0751
Entergy Operations, Inc. - 3 - Director Nuclear Safety Assurance Entergy Operations, Inc.
17265 River Road Killona, LA 70057-0751 General Manager, Plant Operations
Waterford 3 SES
Entergy Operations, Inc.
17265 River Road Killona, LA 70057-0751
Manager, Licensing Entergy Operations, Inc.
17265 River Road Killona, LA 70057-3093
Chairman Louisiana Public Service Commission
P.O. Box 91154
Baton Rouge, LA 70821-9154
Parish President Council
St. Charles Parish P.O. Box 302
Hahnville, LA 70057 Director, Nuclear Safety & Licensing Entergy, Operations, Inc.
440 Hamilton Avenue
White Plains, NY 10601 Louisiana Department of Environmental Quality, Radiological Emergency Planning and Response Division P.O. Box 4312
Baton Rouge, LA 70821-4312 Chief, Technological Hazards Branch FEMA Region VI
800 North Loop 288
Federal Regional Center Denton, TX 76209
Entergy Operations, Inc. - 4 - Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov) Deputy Regional Administrator (Chuck.Casto@nrc.gov)
DRP Director (Dwight.Chamberlain@nrc.gov)
DRP Deputy Director (Anton.Vegel@nrc.gov)
DRS Director (Roy.Caniano@nrc.gov)
DRS Deputy Director (Troy.Pruett@nrc.gov) Senior Resident Inspector (Mark.Haire@nrc.gov) Resident Inspector (Dean.Overland@nrc.gov) Branch Chief, DRP/E (Jeff.Clark@nrc.gov) Senior Project Engineer, DRP/E (Ray.Azua@nrc.gov)
WAT Site Secretary (Linda.Dufrene@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Branch Chief, DRS/TSB (Michael.Hay@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource Regional State Liaison Officer (Bill.Maier@nrc.gov)
NSIR/DPR/EP (Steve.LaVie@nrc.gov)
OEDO RIV Coordinator (Leigh.Trocine@nrc.gov)
ROPreports
File located: R:\_REACTORS\_WAT\2009\WAT 2009004 RP-DHO.doc
SUNSI Rev Compl. Yes No ADAMS Yes No Reviewer Initials
Publicly Avail Yes No Sensitive Yes No Sens. Type Initials
RIV:SRI:DRP/E SPE/DRP/E C:DRS/EB1 C:DRS/EB2 DHOverland
RAzua TRFarnholtz
NFOKeefe /RA/RAzua for
/RA/ /RA/ /RA/ 11/17/09 11/17/09 11/12/09 11/12/09 C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C:DRP/E SGarchow MPShannon
GEWerner JAClark /RA/ /RA/ /RA/ /RA/RAzua
for 11/17/2009
11/13/09 E11/11/09
11/17/09 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION REGION IV
Docket: 05000382 License: NFP-38 Report: 05000382/2009004
Licensee: Entergy Operations, Inc.
Facility: Waterford Steam Electric Station, Unit 3
Location: Hwy. 18 Killona, LA
Dates: July 8, 2009 through October 7, 2009
Inspectors:
D. Overland, Senior Resident Inspector R. Egli, Branch Chief, TTC R. Hickok, Senior Reactor Technology Instructor, TTC
G. Replogle, Senior Project Engineer, RIV M. Chambers, Resident Inspector, Cooper Nuclear Station P. Jayroe, Project Engineer, RIV T. Buchanan, Project Engineer, RIV
S. Anderson, General Engineer, HQ
L. Carson II, Senior Health Physicist
Approved By: Jeff Clark, Chief, Project Branch E Division of Reactor Projects
- 1 - Enclosure
SUMMARY OF FINDINGS IR 05000382/2009004; July 8, 2009 through October 7, 2009; Waterford Steam Electric Station, Unit 3; Operability Evaluations.
The report covered a 3-month period of inspection by resident inspectors and announced baseline inspections by regional based inspectors. One Green noncited violation of significance
was identified. The significance of most findings is indicated by their color (Green, White,
Yellow, or Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing
the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.
A. NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. The inspectors identified a Green non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation. The technical specifications require all four channels (A, B, C, and D) of local power density, departure from nucleate boiling ratio,
and reactor coolant flow instruments to be operable when in Mode 1. These Channel B
instruments require an input from the Channel B log power instrument, which was previously declared inoperable. With the Channel B log power instrument inoperable, the Channel B local power density, departure from nucleate boiling ratio, and reactor
coolant flow instruments should also have been declared inoperable. The licensee
entered this finding in their corrective action program as condition reports CR-WF3-2009-4401 and CR-WF3-2009-4407. The failure to either trip or bypass the inoperable channels within one hour was more than minor because it affected the configuration control attribute of the mitigating systems cornerstone. Specifically, deliberate operator action was required to ensure that proper reactor protection system coincidence and reliability were maintained. Also, if left
uncorrected, the potential existed for Channel B reactor protective trips to be
inadvertently removed while at power. The failure to meet the technical specifications
was considered to be of very low safety significance (Green), since there was no actual loss of safety function. This finding has a cross-cutting aspect in the decision-making
component of the human performance area because the licensee failed to verify the validity of underlying assumptions and identify unintended consequences of failing to
comply with technical specification 3.3.1 by declaring the log power Channel B
inoperable and not placing local power density, departure from nucleate boiling ratio,
and reactor coolant flow instrument channels in either bypass or trip condition (H.1.b). (Section 1R15)
B. Licensee-Identified Violations
None - 2 - Enclosure
REPORT DETAILS Summary of Plant Status The plant began the inspection period on July 8, 2009, at 100 percent power and remained at approximately 100 percent power for the rest of the inspection period. 1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness 1R01 Adverse Weather Protection (71111.01)
.1 Readiness for Impending Adverse Weather Conditions
a. Inspection Scope
Since thunderstorms with potential tornados and high winds were forecast in the vicinity of the facility for October 4, 2009, the inspectors reviewed the licensee's overall preparations/protection for the expected weather conditions. The inspectors evaluated the licensee staff's documented preparations against the site's procedures and
determined that the staff's actions were adequate. During the inspection, the inspectors
focused on plant-specific design features and the licensee's procedures used to respond
to specified adverse weather conditions. The inspector's evaluated operator staffing and
accessibility of controls and indications for
those systems required to control the plant. Additionally, the inspectors reviewed the Updated Final Safety Analysis Report and
verified that operator actions were appropriate as specified by plant-specific procedures. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one readiness for impending adverse weather
condition sample as defined in Inspection Procedure 71111.01-05.
b. Findings No findings of significance were identified. 1R04 Equipment Alignments (71111.04)
.1 Partial Walkdown
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant systems: July 22, 2009: Chemical volume control Train A August 12, 2009: Emergency feedwater Train A August 13, 2009: Low pressure safety injection Train B August 18, 2009: Emergency feedwater Train AB September 15, 2009: High pressure safety injection system Train A
- 3 - Enclosure
The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Final Safety Analysis Report, technical specification requirements, administrative technical specifications, outstanding work orders, condition
reports, and the impact of ongoing work activities on redundant trains of equipment in
order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly
identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of five partial system walkdown samples as defined in Inspection Procedure 71111.04-05.
b. Findings No findings of significance were identified. 1R05 Fire Protection (71111.05)
.1 Quarterly Fire Inspection Tours
a. Inspection Scope
The inspectors conducted fire protection walkdowns that were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas: July 21, 2009: Reactor auxiliary building fire Zones 8B, 8C, 11, and 12 July 22, 2009: Reactor auxiliary building fire Zones 33, 35, 38, and 39 July 30, 2009: Reactor auxiliary building fire Zones 3, 5, and 6 August 3, 2009: Fire Zones Roof E and Roof W August 11, 2009: Reactor auxiliary building fire Zone 16 August 18, 2009: Reactor auxiliary building fire Zones 33, 35, 36, 37, 38, and 39 August 19, 2009: Reactor auxiliary building fire Zone 32 August 20, 2009: Reactor auxiliary building fire Zones 2, Roof E, and Roof W August 23, 2009: Reactor auxiliary building fire Zones 11, 12,13, 8B, and 8C August 24, 2009: Reactor auxiliary building fire Zones 15, 16, 17, 18, 19, 20, and 21 The inspectors reviewed areas to assess if licensee personnel had implemented a fire protection program that adequately controlled combustibles and ignition sources within
- 4 - Enclosure
the plant; effectively maintained fire detection and suppression capability; maintained passive fire protection features in good material condition; and had implemented
adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features, in accordance with the licensee's fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plant's Individual Plant Examination of External Events with later
additional insights, their potential to affect equipment that could initiate or mitigate a plant
transient, or their impact on the plant's ability to respond to a security event. Using the
documents listed in the attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed, that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensee's corrective action program.
Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of ten quarterly fire-protection inspection samples
as defined in Inspection Procedure 71111.05-05.
b. Findings No findings of significance were identified.
.2 Annual Fire Protection Drill Observation (71111.05A) a. Inspection Scope
On September 23, 2009, the inspectors observed a fire brigade activation as the result
of a simulated fire at feed heater drain Pump C. The observation evaluated the
readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee
staff identified deficiencies, openly discussed them in a self-critical manner at the drill debrief, and took appropriate corrective actions. Specific attributes evaluated were: (1) proper wearing of turnout gear and self-contained breathing apparatus; (2) proper
use and layout of fire hoses; (3) employment of appropriate fire fighting techniques; (4) sufficient firefighting equipment brought to the scene; (5) effectiveness of fire brigade
leader communications, command, and control; (6) search for victims and propagation of the fire into other plant areas; (7) smoke removal operations; (8) utilization of pre planned strategies; (9) adherence to the preplanned drill scenario; and (10) drill
objectives. These activities constitute completion of one annual fire-protection inspection sample as defined in Inspection Procedure 71111.05-05.
b. Findings No findings of significance were identified.
- 5 - Enclosure
1R11 Licensed Operator Requalification Program (71111.11) a. Inspection Scope
On August 4, 2009, the inspectors observed a crew of licensed operators in the plant's simulator to verify that operator performance was adequate, evaluators were identifying
and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas: Licensed operator performance Crew's clarity and formality of communications Crew's ability to take timely actions in the conservative direction Crew's prioritization, interpretation, and verification of annunciator alarms Crew's correct use and implementation of abnormal and emergency procedures Control board manipulations Oversight and direction from supervisors Crew's ability to identify and implement appropriate technical specification actions
and emergency plan actions and notifications The inspectors compared the crew's performance in these areas to pre-established operator action expectations and successful critical task completion requirements. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one quarterly licensed-operator requalification program sample as defined in Inspection Procedure 71111.11. b. Findings No findings of significance were identified. 1R12 Maintenance Effectiveness (71111.12Q)
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk
significant systems: August 11, 2009: Seal leakage on chemical volume control charging pumps September 3, 2009: Review of operating experience smart sample FY 2009-01, Inspection of electrical connections for motor control center, circuit breakers and
interfaces
- 6 - Enclosure
The inspectors reviewed events such as where ineffective equipment maintenance has resulted in valid or invalid automatic actuations of engineered safeguards systems and
independently verified the licensee's actions to address system performance or condition
problems in terms of the following: Implementing appropriate work practices Identifying and addressing common cause failures Scoping of systems in accordance with 10 CFR 50.65(b) Characterizing system reliability issues for performance Charging unavailability for performance Trending key parameters for condition monitoring Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or (a)(2) Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance
through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as
requiring the establishment of appropriate and adequate goals and corrective
actions for systems classified as not having adequate performance, as described
in 10 CFR 50.65(a)(1) The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of two quarterly maintenance effectiveness samples as defined in Inspection Procedure 71111.12-05.
b. Findings No findings of significance were identified. 1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13) a. Inspection Scope
The inspectors reviewed licensee personnel's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were
performed prior to removing equipment for work: July 29, 2009: Scheduled elective maintenance outage for containment fan coolers Train B to calibrate containment fan cooler Header B CCW return temperature control valve solenoid Valve CC-835B
- 7 - Enclosure
August 3, 2009: Scheduled surveillance of reactor protection system Channel A September 9, 2009: Scheduled activity to remove high pressure safety injection Pump AB from high pressure safety injection Train A alignment and align high pressure safety injection Pump A to Train A September 11, 2009: Emergent maintenance to replace station Battery AB, Cell 31 with a spare cell due to degraded cell voltage The inspectors selected these activities based on potential risk significance relative to
the reactor safety cornerstones. As applicable for each activity, the inspectors verified
that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four maintenance risk assessments and emergent work control inspection samples as defined in Inspection
Procedure 71111.13-05.
b. Findings No findings of significance were identified. 1R15 Operability Evaluations (71111.15) a. Inspection Scope
The inspectors reviewed the following issues: July 14, 2009: Low individual cell voltage on vital 125 vdc station Battery AB
Cell 39 August 11, 2009: Unplanned load variations during emergency diesel generator
Train A surveillance
August 12, 2009: Emergency diesel generator Train A Relay EG EREL 2342(J) found out of calibration during surveillance
August 20, 2009: Channel B local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments, when Channel B log power
instrument was inoperable
- 8 - Enclosure
The inspectors selected these potential operability issues based on the risk-significance of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that technical specification operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and
design criteria in the appropriate sections of the technical specifications and Updated
Safety Analysis Report to the licensee's evaluations, to determine whether the components or systems were operable. Where compensatory measures were required
to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with
operability evaluations. Specific documents reviewed during this inspection are listed in
the attachment. These activities constitute completion of four operability evaluations inspection samples as defined in Inspection Procedure 71111.15-05
b. Findings Introduction: The inspectors identified a Green non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation. The technical specifications
require all four channels (A, B, C, and D) of local power density, departure from nucleate
boiling ratio, and reactor coolant flow instruments to be operable when in Mode 1.
These Channel B instruments require an input from the Channel B log power instrument,
which was previously declared inoperable. With the Channel B log power instrument
inoperable, the Channel B local power density, departure from nucleate boiling ratio, and
reactor coolant flow instruments should also have been declared inoperable.
Description: On Aug 20, 2009, the inspector observed the performance of procedure MI-003-126, Revision 14, "Core Protection Calculator Functional." During the
performance of the test procedure, the inspector noted that CPC Channel B high log
power trip was bypassed. The inspector asked why technical specification 3.3.1 had not
been entered due to the inoperable log power Channel B instrument. Technical
specification 3.3.1, Reactor Protective Instrumentation, requires that the reactor protective instrumentation channels and bypasses contained in Table 3.3-1 be operable in accordance with the requirements of the table. Table 3.3-1 requires all four channels
of local power density (LPD), departure from nucleate boiling ratio (DNBR), and reactor coolant flow instruments to be operable in Mode 1.
Log power Channel B provides a high log power automatic bypass removal signal for LPD, DNBR, and reactor coolant flow instrumentation channels. Technical specification 3.3.1, Table 3.3-1 requires the high log power bypass shall be
automatically removed when reactor power is greater than or equal to
10-4% of rated thermal power. When in Mode 1, reactor power is greater than 10
-4% of rated thermal power. The inspectors determined that when a log power instrument is out of service, the automatic removal of the high log power bypass function is inoperable and thus the associated protective channels of LPD, DNBR, and reactor coolant flow are also inoperable.
- 9 - Enclosure
The log power Channel B instrument was originally declared inoperable on Sept 1, 2008. The operability determination concluded that since the plant was in Mode 4, only two log
power channels were required, therefore entry into technical specification 3.3.1 was not required. On Sept 9, 2008, the plant entered Mode 2 with log power Channel B still inoperable. The operability was not revised to reflect the change in plant conditions. In
accordance with technical specification 3.3.1, operators should have taken action to
place the associated LPD, DNBR, and reactor coolant flow protective channels to either bypass or trip within one hour. On Aug 22, 2009, after considering the inspector's question, the licensee declared LPD Channel B and DNBR Channel B inoperable, and placed both instruments in bypass. During a subsequent control room tour, the inspector verified that LPD and DNBR were bypassed, however noticed that reactor coolant flow Channel B had not been bypassed.
The inspector asked the shift manager if technical specification 3.3.1, Table 3.3-1
notation (C) affected any other trips. Upon further assessment, operations personnel
determined that reactor coolant low flow was also affected and declared steam generator flow Channel B to be inoperable, as well.
Analysis: The failure to either trip or bypass the inoperable channels within one hour was more than minor because it affected the configuration control attribute of the mitigating systems cornerstone. Specifically, deliberate operator action was required to ensure that proper reactor protection system coincidence and reliability were maintained. Also, if left uncorrected, the potential existed for Channel B reactor protective trips to be
inadvertently removed while at power. The failure to meet the technical specifications
was considered to be of very low safety significance (Green), since there was no actual loss of safety function. This finding has a cross-cutting aspect in the decision-making
component of the human performance area because the licensee failed to verify the validity of underlying assumptions and identify unintended consequences of failing to
comply with technical specification 3.3.1 by declaring the log power Channel B
inoperable and not placing DNBR, LPD, and reactor coolant flow channels in either
bypass or trip condition (H.1.b).
Enforcement: Technical specification 3.3.1, "Reactor Protective Instrumentation," requires all four channels of LPD, DNBR, and reactor coolant flow to be operable and
able to have the high log power bypass autom
atically removed when reactor power is
greater than or equal to 10
-4% percent of rated thermal power. Contrary to this, on September 9, 2008, the licensee did not comply with the limiting condition for operation action statement for technical specification 3.3.1 which states, "the inoperable channel is
placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />." The plant remained in this condition until August 22, 2009. This violation has been determined to be of very low safety significance and was entered into their corrective action program in condition
reports CR-WF3-2009-4401 and CR-WF3-2009-4407. Therefore, this violation is being
treated as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC
- 10 - Enclosure
1R18 Plant Modifications (71111.18) a. Inspection Scope
The inspectors reviewed the following temporary/permanent modifications to verify that the safety functions of important safety systems were not degraded: August 26, 2009: Permanent modification of containment vacuum relief valves such that once the valves are automatically opened, they remain open until
manually closed.
August 7, 2009: Temporary modification to revise the setpoint for the reactor coolant Pump 2A upper thrust bearing high temperature alarm to reduce
nuisance alarms in the control room.
September 14, 2009: Temporary modification to replace station Battery AB, Cell 31 with a new cell. The old Cell 31 was left in place and jumpered around, while the new Cell 31 was installed at the end of the battery rack.
The inspectors reviewed the temporary modification and the associated safety evaluation screening against the system design bases documentation, including the
Updated Final Safety Analysis Report and the technical specifications, and verified that
the modification did not adversely affect the system operability/availability. The inspectors also verified that the installation and restoration were consistent with the
modification documents and that configuration control was adequate. Additionally, the
inspectors verified that the temporary modification was identified on control room
drawings, appropriate tags were placed on the affected equipment, and licensee
personnel evaluated the combined effects on mitigating systems and the integrity of
radiological barriers. The inspectors reviewed key affected parameters associated with energy needs, materials/replacement components, timing, heat removal, control signals, equipment protection from hazards, operations, flow paths, pressure boundary, ventilation
boundary, structural, process medium properties, licensing basis, and failure modes for the modification listed below. The inspectors verified that modification preparation,
staging, and implementation did not impair emergency/abnormal operating procedure actions, key safety functions, or operator response to loss of key safety functions; postmodification testing will maintain the plant in a safe configuration during testing by
verifying that unintended system interactions will not occur, systems, structures and
components' performance characteristics still meet the design basis, the appropriateness of modification design assumptions, and the modification test
acceptance criteria will be met; and licensee personnel identified and implemented appropriate corrective actions associated with permanent plant modifications. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of three samples for temporary and permanent plant modifications as defined in Inspection Procedure 71111.18-05
- 11 - Enclosure
b. Findings No findings of significance were identified. 1R19 Postmaintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed the following postmaintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability: June 23, 2009: Replacement of high pressure safety injection Pump B Tyco time delay relay following the failure of the relay to start the pump during a routine
surveillance test
July 23, 2009: Replacement of seal package on chemical volume control charging Pump B to reduce reactor coolant system unidentified leakage
July 29, 2009: Scheduled elective maintenance calibration of containment fan cooler Header B CCW return temperature control valve solenoid Valve CC-835B
August 4, 2009: Corrective maintenance to repair the actuator for steam generator SG1 main steam atmospheric dump Valve MS-116A
August 11, 2009: Scheduled preventative maintenance to clean, inspect, and test emergency diesel generator Train A Relay EG EREL 2342(J)
September 9, 2009: Scheduled preventative maintenance to replace the pulsation dampener and perform motor maintenance on chemical volume control charging Pump AB
September 14, 2009: Emergent maintenance to replace station Battery AB, Cell 31 with a spare cell, due to degraded voltage on the cell
September 29, 2009: Scheduled preventative maintenance to check the overcurrent trip on the breaker for non-nuclear safety return header isolation Valve CC-562.
The inspectors selected these activities based upon the structure, system, or
component's ability to affect risk. The inspectors evaluated these activities for the following (as applicable): The effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed
Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate
- 12 - Enclosure
The inspectors evaluated the activities against the technical specifications, the Updated Final Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and
various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with postmaintenance tests
to determine whether the licensee was identifying problems and entering them in the
corrective action program and that the problems were being corrected commensurate
with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of eight postmaintenance testing inspection sample(s) as defined in Inspection Procedure 71111.19-05.
b. Findings No findings of significance were identified. 1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the Updated Final Safety Analysis Report, procedure requirements, and technical specifications to ensure that the six surveillance activities
listed below demonstrated that the systems, structures, and/or components tested were
capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the significant surveillance test attributes were adequate to address the following:
Preconditioning Evaluation of testing impact on the plant Acceptance criteria Test equipment Procedures Jumper/lifted lead controls Test data Testing frequency and method demonstrated technical specification operability Test equipment removal Restoration of plant systems Fulfillment of ASME Code requirements
- 13 - Enclosure
Updating of performance indicator data Engineering evaluations, root causes, and bases for returning tested systems, structures, and components not meeting the test acceptance criteria were correct
Reference setting data Annunciators and alarms setpoints The inspectors also verified that licensee personnel identified and implemented any needed corrective actions associated with the surveillance testing.
August 6, 2009: Safety related electrical Bus 3A undervoltage relay calibration August 10, 2009: Emergency diesel generator Train A surveillance August 20, 2009: Core protection calculator Train B surveillance August 22, 2009: Plant protection system Channel B surveillance August 24, 2009: Emergency diesel generator and subgroup relays Train B September 14, 2009: High pressure safety injection Train AB
Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of six surveillance testing inspection samples as
defined in Inspection Procedure 71111.22-05.
b. Findings No findings of significance were identified. Cornerstone: Emergency Preparedness 1EP6 Drill Evaluation (71114.06) .1 Training Observations
a. Inspection Scope The inspectors observed a training evolution for licensed operators on September 17, 2009, which required emergency plan implementation by a licensee operations crew. This evolution was planned to be evaluated and included in performance indicator data regarding drill and exercise performance. The inspectors observed event classification and notification activities performed by the crew. The inspectors also attended the
postevolution critique for the scenario. The focus of the inspectors' activities was to note
any weaknesses and deficiencies in the crew's performance and ensure that the
licensee evaluators noted the same issues and entered them into the corrective action
program. As part of the inspection, the inspectors reviewed the scenario package and
other documents listed in the attachment. These activities constitute completion of one sample as defined in Inspection Procedure 71114.06-05.
- 14 - Enclosure
b. Findings No findings of significance were identified. 2. RADIATION SAFETY
Cornerstone: Occupational and Public Radiation Safety
2OS2 ALARA Planning and Controls (71121.02)
a. Inspection Scope
The inspector assessed licensee performance with respect to maintaining individual and
collective radiation exposures ALARA. The inspector used the requirements in 10 CFR Part 20 and the 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 associated work activity exposure estimates which were likely to result in the
highest personnel collective exposures
- Site-specific trends in collective exposures, plant historical data, and source-term measurements
- Five work activities of highest exposure significance completed during the last
outage
- ALARA work activity evaluations, exposure estimates, and exposure mitigation
requirements
- Intended versus actual work activity doses and the reasons for any
inconsistencies
- Person-hour estimates provided by maintenance planning and other groups to the radiation protection group with the actual work activity time requirements
- Post-job (work activity) reviews
- Assumptions and basis for the current annual collective exposure estimate, the methodology for estimating work activity exposures, the intended dose outcome,
and the accuracy of dose rate and man-hour estimates
- Method for adjusting exposure estimates, or re-planning work, when unexpected changes in scope or emergent work were encountered
- Exposure tracking system
- 15 - Enclosure
- Exposures of individuals from selected work groups
- Records detailing the historical trends and current status of tracked plant source terms and contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry
- Declared pregnant workers during the current assessment period, monitoring controls, and the exposure results
- Self-assessments, audits, and special reports related to the ALARA program since the last inspection
- Resolution through the corrective action process of problems identified through post-job reviews and post-outage ALARA report critiques
The inspector completed 11 of the required 15 samples and 5 of the optional samples as
defined in IP 71121.02-05.
4. OTHER ACTIVITIES 4OA1 Performance Indicator Verification (71151)
.1 Data Submission Issue
a. Inspection Scope
The inspectors performed a review of the data submitted by the licensee for the second quarter of 2009 performance indicators for any obvious inconsistencies prior to its public
release in accordance with Inspection Manual Chapter 0608, "Performance Indicator
Program." This review was performed as part of the inspectors' normal plant status activities and, as such, did not constitute a separate inspection sample.
b. Findings No findings of significance were identified.
.2 Safety System Functional Failures
a. Inspection Scope
The inspectors sampled licensee submittals for the Safety System Functional Failures performance indicator for the period from the second quarter of 2008 through the second
quarter of 2009. To determine the accuracy of the performance indicator data reported
during those periods, performance indicator definitions and guidance contained in NEI
Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 5,
and NUREG-1022, "Event Reporting Guidelines 10 CFR 50.72 and 50.73" definitions and guidance were used. The inspectors reviewed the licensee's operator narrative logs, operability assessments, maintenance rule records, maintenance work orders, - 16 - Enclosure
issue reports, event reports and NRC Integrated Inspection reports for the period beginning the second quarter of 2008 through the second quarter of 2009 to validate the
accuracy of the submittals. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report. These activities constitute completion of one safety system functional failures sample as defined in Inspection Procedure 71151-05.
b. Findings No findings of significance were identified.
.3 Mitigating Systems Performance I
ndex - Emergency ac Power System
a. Inspection Scope
The inspectors sampled licensee submittals for the Mitigating Systems Performance Index (MSPI) - Emergency ac Power System performance for the period from the second quarter of 2008 through the second quarter of 2009. To determine the accuracy
of the performance indicator data reported during those periods, performance indicator
definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment
Performance Indicator Guideline," Revision 5, was used. The inspectors reviewed the
licensee's operator narrative logs, miti
gating systems performance index derivation reports, issue reports, event reports and NRC integrated inspection reports for the period beginning the second quarter of 2008 through the second quarter of 2009 to validate the
accuracy of the submittals. The inspectors reviewed the mitigating systems performance index component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensee's issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report. These activities constitute completion of one mitigating systems performance index emergency ac power system sample as defined in Inspection Procedure 71151-05.
b. Findings No findings of significance were identified.
.4 Mitigating Systems Performance Index - Cooling Water Systems
a. Inspection Scope
The inspectors sampled licensee submittals for the Mitigating Systems Performance Index - Cooling Water Systems performance for the period from the second quarter of 2008 through the second quarter of 2009. To determine the accuracy of the
performance indicator data reported during those periods, performance indicator
definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment
- 17 - Enclosure
Performance Indicator Guideline," Revision 5, was used. The inspectors reviewed the licensee's operator narrative logs, issue reports, mitigating systems performance index derivation reports, event reports and NRC integrated inspection reports for the period beginning the second quarter of 2008 through the second quarter of 2009 to validate the accuracy of the submittals. The inspectors reviewed the mitigating systems performance
index component risk coefficient to determine if it had changed by more than 25 percent
in value since the previous inspection, and if so, that the change was in accordance with
applicable NEI guidance. The inspectors also reviewed the licensee's issue report
database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report. These activities constitute completion of one mitigating systems performance index cooling water system sample as defined in Inspection Procedure 71151-05.
b. Findings No findings of significance were identified.
.16 Occupational Exposure Control Effectiveness (OR01)
a. Inspection Scope
The inspector sampled licensee submittals for the Occupational Radiological
Occurrences performance indicator for the first quarter of 2009 through the
third quarter of 2009. To determine the accuracy of the performance indicator data reported during
those periods, performance indicator definitions and guidance contained in NEI
Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 5,
were used. The inspector reviewed the licensee's assessment of the performance
indicator for occupational radiation safety to determine if indicator related data was
adequately assessed and reported. To assess the adequacy of the licensee's performance indicator data collection and analyses, the inspector discussed with radiation protection staff, the scope and breadth of its data review, and the results of
those reviews. The inspector independently reviewed electronic dosimetry dose rate
and accumulated dose alarm and dose reports and the dose assignments for any intakes that occurred during the time period reviewed to determine if there were potentially unrecognized occurrences. The inspector also conducted walkdowns of numerous locked high and very high radiation area entrances to determine the adequacy
of the controls in place for these areas.
These activities constitute completion of the occupational radiological occurrences
sample as defined in Inspection Procedure 71151-05.
b. Findings
No findings of significance were identified.
- 18 - Enclosure
.17 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (PR01)
a. Inspection Scope
The inspector sampled licensee submittals for the Radiological Effluent Technical
Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences
performance indicator for the first quarter of 2009 through the
third quarter of 2009. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 5, was used. The
inspector reviewed the licensee's issue report database and selected individual reports
generated since this indicator was last reviewed to identify any potential occurrences
such as unmonitored, uncontrolled, or improperly calculated effluent releases that may have impacted offsite dose. The inspector reviewed gaseous effluent summary data and the results of associated offsite dose calculations for selected dates during the third
quarter of 2009 to determine if indicator results were accurately reported. The inspector
also reviewed the licensee's methods for quantifying gaseous and liquid effluents and determining effluent dose. Additionally, the inspector reviewed the licensee's historical 10 CFR Part 50.75(g) file and selectively reviewed the licensee's analysis for discharge pathways resulting from a spill, leak, or unexpected liquid discharge focusing on those
incidents which occurred over the last few years.
These activities constitute completion of the radiological effluent technical
specifications/offsite dose calculation manual radiological effluent occurrences sample as defined in Inspection Procedure 71151-05.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152) Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical
Protection
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensee's
corrective action program at an appropriate threshold, that adequate attention was being
given to timely corrective actions, and that adverse trends were identified and
addressed. The inspectors reviewed attributes that included: the complete and accurate identification of the problem; the timely correction, commensurate with the safety significance; the evaluation and disposition of performance issues, generic
implications, common causes, contributing factors, root causes, extent of condition
- 19 - Enclosure
reviews, and previous occurrences reviews; and the classification, prioritization, focus, and timeliness of corrective. Minor issues entered into the licensee's corrective action
program because of the inspectors' observations are included in the attached list of
documents reviewed. These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure, they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.
b. Findings No findings of significance were identified.
.2 Daily Corrective Action Program Reviews
a. Inspection Scope
In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensee's corrective action program. The inspectors accomplished this through review of the station's daily corrective action documents. The inspectors performed these daily reviews as part of their daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.
b. Findings No findings of significance were identified.
.3 Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a review of the licensee's corrective action program and associated documents to identify trends that could indicate the existence of a more
significant safety issue. The inspectors focused their review on repetitive equipment issues, but also considered the results of daily corrective action item screening discussed in Section 4OA2.2, above, licensee trending efforts, and licensee human
performance results. The inspectors nominally considered the 6-month period of
January 2009 through July 2009, for a review of Operating Experience Smart Sample:
OpESS FY2009-02, "A Negative trend and Recurring Events Involving feedwater systems" as it applies to t
he emergency feedwater system. The inspectors also included issues documented outside the normal corrective action program in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance
audit/surveillance reports, self-assessment reports, and Maintenance Rule assessments.
The inspectors compared and contrasted their results with the results contained in the
licensee's corrective action program trending reports. Corrective actions associated with
- 20 - Enclosure
- 21 - Enclosure a sample of the issues identified in the licensee's trending reports were reviewed for adequacy. These activities constitute completion of one single semi-annual trend inspection sample as defined in Inspection Procedure 71152-05.
b. Findings No findings of significance were identified. 4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel and Activities
a. Inspection Scope
During the inspection period, the inspectors performed observations of security force personnel and activities to ensure that the activities were consistent with Waterford Steam Electric Station security procedures and regulatory requirements relating to
nuclear plant security. These observations took place during both normal and off-normal plant working hours. These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status review and inspection activities.
b. Findings No findings of significance were identified. 4OA6 Meetings Exit Meeting Summary
On September 18, 2009, the team presented the inspection results to Mr. J. Kowalewski,
Vice President, Operations, and other members of his staff who acknowledged the findings.
The team confirmed that proprietary information was not provided or examined during the inspection.
On October 1, 2009, the inspectors presented the inspection results to Mr. Joe Kowalewski, and
other members of the licensee staff. The licensee acknowledged the issues presented. The
inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
M. Adams, Supervisor, System Engineering S. Anders, Manager, Plant Security
C. Arnone, Plant Manager
J. Brawley, ALARA Supervisor, Radiation Protection
B. Briner, Technical Specialist IV, Componet Engineering
K. Christian, Director, Nuclear Safety Assurance K. Cook, Manager, Operations L. Dauzat, Supervisor, Radiation Protection
D. Dufrene; Technician, Radiation Protection
C. Fugate, Assistant Manager, Operations M. Haydel, Supervisor, Programs and Components J. Kowalewski, Vice President of Operations J. Lewis, Manager, Emergency Preparedness
B. Lindsey, Manager, Maintenance
M. Mason, Senior Licensing Specialist, Licensing
W. McKinney, Manager, Corrective Action and Assessments C. Miller, Lead Supervisor, Radiation Protection R. Murillo, Manager, Licensing
K. Nicholas, Director, Engineering
B. Piluti, Manager, Radiation Protection
J. Polluck, Engineer, Licensing
R. Putnam, Manager, Programs and Components S. Ramzy; Specialist, Radiation Protection J. Ridge, Manager, Quality Assurance
J. Solaski, Quality Assurance Auditor J. Williams, Senior Licensing Specialist, Licensing NRC Personnel
S. Anderson, General Engineer, HQ T. Buchanan, Project Engineer, RIV
L. Carson II, Senior Health Physicist
M. Chambers, Resident Inspector, Cooper Nuclear Station R. Egli, Branch Chief, Technical Training Center R. Hickok, Senior Reactor Technology Instructor, Technical Training Center
P. Jayroe, Project Engineer, RIV
G. Replogle, Senior Project Engineer, RIV
A-1 Attachment
LIST OF ITEMS OPENED AND CLOSED
Opened and Closed
NCV Failure to Follow Technical Specification Requirements for Reactor Protective Instrumentation
LIST OF DOCUMENTS REVIEWED Section 1RO1: Adverse Weather Protection
CONDITION REPORTS
PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION OP-901-521 Sever Weather and Flooding
301 Section 1RO4: Equipment Alignment
CONDITION REPORTS
CR-WF3-2009-0607 CR-WF3-2009-0737 CR-WF3-2009-1189 CR-WF3-2009-1624 CR-WF3-2009-2869
WORK ORDERS
190714 PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION OP-903-045 Emergency Feedwater Flow Path Lineup Verification
5 OP-009-008
Safety Injection System
26 OP-002-005 Chemical and Volume Control
28 SD-CVC Chemical and Volume Control System Description
6 SD-SI Safety Injection System Description
6 A-2 Attachment
Section 1RO5: Fire Protection CONDITION REPORTS
CR-WF3-2009-04034 CR-WF3-2009-04035 CR-WF3-2009-04060
PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION UNT-005-013
10 OP-009-004 Fire Protection
305 MM-004-424 Building Fire Hose Station Inspection and Hose
Replacement
10 MM-007-010 Fire Extinguisher Inspection and Extinguisher Replacement
302 FP-001-014 Duties of a Fire Watch
14 FP-001-015 Fire Protection Impairments
302 DBD-018 Appendix R/Fire Protection
FP-001-015 Fire Protection Impairments
302 FP-001-018 Pre-fire Plan Strategies, Development, And Revision
300 UNT-007-006
Housekeeping
301 EN-DC-161
Control of Combustibles
003 UNT-007-060
Control of Loose Items
302 UNT-005-013
010 Engineering Calculations F91-044
01 Engineering Calculations F91-019
0 Section 1R11: Licensed Operator Requalification Program
PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION Simulator Scenario Number E-70
Simulator Scenario Number E-125
OP-901-201 Steam Generator Level Control System Malfunction
009 A-3 Attachment
OP-902-000 Standard Post Trip Actions
010 OP-902-008
Safety function Recovery Procedure
015 OP-901-110 Pressurizer Level Control Malfunction
005 OP-901-311 Loss of Train B Safety Bus
302 OP-901-102 CEA or CEDMCS Malfunciton
300 OP-902-001 Reactor Trip Recovery 011 OP-902-002 Loss of Coolant Accident Recovery Procedure
012 Section 1R12: Maintenance Effectiveness CONDITION REPORTS
CR-WF3-2007-3497 CR-WF3-2008-4306 CR-WF3-2008-3836 CR-WF3-2009-0506 CR-WF3-2008-4189 CR-WF3-2008-4611 CR-WF3-2009-1190 CR-WF3-2009-4131 CR-WF3-2008-4297 CR-WF3-2008-4765 CR-WF3-2009-2862 CR-WF3-2009-3810
CR-WF3-2008-1072 CR-WF3-2008-2410 CR-WF3-2008-2352 CR-WF3-2008-4332 CR-WF3-2008-1796 CR-WF3-2008-2810 CR-WF3-2008-2579 CR-WF3-2008-5045 CR-WF3-2008-1807 CR-WF3-2008-3363 CR-WF3-2008-4127 CR-WF3-2008-5273
CR-WF3-2008-2066 CR-WF3-2008-2346 CR-WF3-2008-4173 CR-WF3-2009-0955 CR-WF3-2009-1200 CR-WF3-2009-1284 CR-WF3-2009-4015 CR-WF3-2009-4324
PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION EN-DC-206 Maintenance Rule
1 NUMARC 93-01 Industry Guideline for Monitoring the Effectiveness of maintenance at Nuclear Power Plants
3 Section 1R13: Maintenance Risk Assessment and Emergent Work Controls
WORK ORDERS
51802942 52039753 0019397401
52192184 197692 PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION OI-037-000 Operations Risk Management Guideline
300 EN-WM-101 On-Line Work Management Process
1 A-4 Attachment
W2.502 Configuration risk Management Program
000 OP-100-010 Equipment Out of Service
303 OP-903-107 Plant Protection System channel A & B & C & D Functional Test
303 OP-903-030 Safety Injection Pump Operability Verification
18 OP-009-008
Safety Injection System
26 OP-006-003 125 VDC Electrical Distribution
301 ME-003-200 Station Battery Bank and Charger (Weekly)
301 ME-003-210 Station Battery Bank and Charger (Quarterly)
12 Section 1R15: Operability Evaluations
CONDITION REPORTS
CR-WF3-2009-4466 CR-WF3-2009-4163 CR-WF3-2009-4395 CR-WF3-2009-4401 CR-WF3-2009-4407 CR-WF3-2009-3540 CR-WF3-2009-4139 CR-WF3-2009-3448 CR-WF3-2009-3557
WORK ORDERS
5180191 52038533 PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION EN-OP-104 Operability Determinations
4 ME-003-200 Station Battery Bank and Charger (Weekly)
301 ME-003-210 Station Battery Bank and Charger (Quarterly)
12 OP-006-003
125 Vdc Electrical Distribution
301 OP-006-001 Plant Distribution System
305 MI-003-126 Core Protection Calculator Functional
14 SD-PPS Plant Protection System Description
0 OP-903-107 Plant Protection System Channel A, B, C, D, Functional Test
303 TSTF-324 Correct logarithmic power vs. RTP
1 A-5 Attachment
Section 1R18: Plant Modifications
CONDITION REPORTS
WORK ORDERS
203111 197692 PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION / DATE EN-DC-136 Temporary Modifications
4 EC NO: 706 Modification of containment relief valves
2/3/2007 EN-WM-105 Implement EC 15451
2/3/2007 ME-004-213 Battery Intercell Connections
14 16496 Temporary Modification
Section 1R19: Postmaintenance Testing
CONDITION REPORTS
CR-WF3-2009-3102 CR-WF3-2009-4304 CR-WF3-2009-3448 CR-WF3-2009-4139 CR-WF3-2009-4766
WORK ORDERS
199029 51802942 52039753 0019397401
199977 188048 52040097 52038057 51523543 201698 52194563 197692 5180191 PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION OP-903-030
Safety Injection Pump Operability Verification
15 OP-903-068 Emergency Diesel Generator Operability and Subgroup Relay Operability Verification
303 OP-009-008
Safety Injection System
25 A-6 Attachment
Section 1R19: Postmaintenance Testing
CONDITION REPORTS
CR-WF3-2009-3102 CR-WF3-2009-4304 CR-WF3-2009-3448 CR-WF3-2009-4139 CR-WF3-2009-4766 WORK ORDERS
199029 51802942 52039753 0019397401
199977 188048 52040097 52038057
51523543 201698 52194563 197692 5180191 PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION OP-903-118 Primary Auxiliaries Quarterly IST Valve Tests
18 OP-903-037 Containment Cooling Fan Operability Verification
5 OP-903-119 Secondary Auxiliaries Quarterly IST Valve Tests
9 OP-903-120 Containment and Miscellaneous Systems Quarterly IST Valve Tests
9 OP-903-003 Charging Pump Operability Check
301 ME-004-213 Battery Intercell Connections
14 OP-903-118 Primary Auxiliaries Quarterly IST Valve Tests
18 ME-007-002
15 SD-CC Component Cooling Water and Auxiliary Component
Cooling Water System Description
7 STA-001-005 Leakage testing of Air and Nitrogen Accumulators for Safety Related Valves
304 Section 1R22: Surveillance Testing
CONDITION REPORTS
CR-WF3-2009-04053 CR-WF3-2009-04072 CR-WF3-2009-04073 CR-WR3-2009-4395 CR-WF3-2009-4401 CR-WF3-2009-4203 CR-WF3-2008-4163 CR-WR3-2009-4466
A-7 Attachment
PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION ME-003-318 G.E. Undervoltage Relay Model 121AV55C
303 OP-009-002 Emergency Diesel Generator Start Evaluation [Data Sheet]
310 OP-009-002 Diesel Generator Start Running Log
310 OP-903-068 Emergency Diesel Generator A Surveillance Test
OP-903-068 Emergency Diesel Generator and Subgroup Relay Operability Verification - Train B
303 OP-903-030 Safety Injection Pump Operability Verification
18 OP-009-008
Safety Injection System
26 OP-903-107 Plant Protection System Channel B Functional Test
303 MI-003-126 Core Protection Calculator Functional
014 Section 1EP6: Drill Evaluation
PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION EP-001-001 Recognition and Classification of Emergency Conditions
22 EP-001-030 Site Area Emergency
300 EP-001-040 General Emergency
300 Scenario DEP 2007-02
Section 2OS2: ALARA Planning and Controls
PROCEDURES/DOCUMENTS
NUMBER TITLEREVISION EN-RP-102 Radiological Control
0 EN-RP-105 Radiation Work Permits
4 EN-RP-106 Radiological Survey Documentation
2 EN-RP-141 Job Coverage
6 EN-DIR-RP-002 Radiation Protection Performance Indicator
0 A-8 Attachment
AUDITS, SELF-ASSESSMENTS, AND SURVEILLANCES
NUMBER TITLE DATE QA-14/15-2009-WF3-1 Radiation Protection/Radwaste Audit
Quality Oversight Observations
May 2008
RADIATION WORK PERMITS
RWP# RWP DESCRIPTION
2008-0511 1R15 S/G Primary Side Eddy Current Testing Inspection and Repair
2008-0610 1R Scaffolding
2008-0631 1R15 Alloy 600 Mitigation Activities Pressurizer/Hot Legs (Weld Overlay)
2008-0702 Reactor Disassembly
2008-0705 Reactor Reassembly
CONDITION REPORTS
CR-WF3-2008-1699 CR-WF3-2008-1776 CR-WF3-2008-1793 CR-WF3-2008-1946 CR-WF3-2008-1989 CR-WF3-2008-2027 CR-WF3-2008-2347 CR-WF3-2008-4495 CR-WF3-2009-4959 CR-WF3-2009-4969
MISCELLANEOUS
TITLE DATE Waterford 3 Refuel Reactor Coolant System Dose Equivalent Iodine September 10, 2009
Reactor Coolant System Cleanup Flow Chart
5-Year ALARA Plan
Refueling Outage 15 Report Failed Fuel Shutdown Mitigation Plan
Section 4OA1: Performance Indicator Verification
PROCEDURES
NUMBER TITLEREVISION NEI 99-02 Regulatory Assessment Performance Indicator Guideline
5 EN-LI-114 Performance Indicator Process
4 A-9 Attachment
A-10 Attachment
Section 4OA2: Identification and Resolution of Problems CONDITION REPORTS
CR-WF3-2008-4000 CR-WF3-2008-4748 CR-WF3-2008-5793 CR-WF3-2009-0089 CR-WF3-2009-0570 CR-WF3-2009-1416 CR-WF3-2009-2604 CR-WF3-2009-3294 CR-WF3-2009-0754 CR-WF3-2009-1446 CR-WF3-2009-2706 CR-WF3-2009-3651 CR-WF3-2009-0770 CR-WF3-2009-2136 CR-WF3-2009-
WORK ORDERS
178225 51665138 PROCEDURES
NUMBER TITLE REVISION EFW System Health Report 1
st Quarter 2009
4/30/09