IR 05000219/2012007
| ML120790052 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/16/2012 |
| From: | Doerflein L T Engineering Region 1 Branch 2 |
| To: | Pacilio M J Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-12-007 | |
| Download: ML120790052 (22) | |
Text
ffi UNITED STATES NUCLEAR REGULATORY COIUIUI ISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PENNSYLVANIA 19406-1415 l.larch 76, 2AL2 Mr. MichaelJ.
Pacilio Senior Vice President, Exelon Generation Company, LLC President and Chief Nuclear Officer, Exelon Nuclear 4300 Winfield Road Warenville.
lL 60555
SUBJECT: OYSTER CREEK GENERATING STATION _ NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODI FI CATI ONS TEAM I N SPECTI ON REPORT O5OOO2 1 91201 2OO7
Dear Mr. Pacilio:
On February 10, 2012, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Oyster Creek Generating Station. The enclosed inspection report documents the inspection results, which were discussed on February 10,2012, with Mr. M. Massaro, Site Vice President, and other members of your staff.The inspection 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.ln conducting the inspection, the team reviewed selected procedures, calculations and records, observed activities, and interviewed station personnel.
Based on the results of this inspection, no findings were identified.
ln accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system, Agencywide Documents Access and Management 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,Engineering Branch 2 Division of Reactor Safety Docket No. 50-219 License No. DPR-16
Enclosure:
f nspection Report 050002 1 912012007 M
Attachment:
Supplemental Information cc Mencl: Distribution via ListServ
SUMMARY OF FINDINGS
lR 0500021912012007;0112312012 - 0211012012;
Oyster Creek Generating Station; Engineering Specialist Plant Modifications Inspection.
This report covers a two week on-site inspection period of the evaluations of changes, tests, or experiments and permanent plant modifications.
The inspection was conducted by three region based engineering inspectors.
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.No findings were identified.
REPORT DETAILS
1. REACTORSAFETY Cornerstones:
Initiating Events, Mitigating Systems, and Barrier Integrity 1R17 Evaluations of Chanses. Tests, or Experiments and Permanent Plant Modifications (tP 71111.17).1 Evaluations of Chanqes. Tests. or Experiments (25 samples)a. Inspection Scope The team reviewed five safety evaluations to determine whether the changes to the facility or procedures, as described in the Updated Final Safety Analysis Report (UFSAR), had been reviewed and documented in accordance with 10 CFR 50.59 requirements.
In addition, the team evaluated whether Exelon had been required to obtain NRC approval prior to implementing the changes. The team interviewed plant staff and reviewed supporting information including calculations, analyses, design change documentation, procedures, the UFSAR, the Technical Specifications (TS), and plant drawings to assess the adequacy of the safety evaluations.
The team compared the safety evaluations and supporting documents to the guidance and methods provided in Nuclear Energy Institute (NEl) 96-07, "Guidelines for 10 CFR 50.59 Evaluations," as endorsed by NRC Regulatory Guide 1.187 , "Guidance for lmplementation of 10 CFR 50.59, Changes, Tests, and Experiments," to determine the adequacy of the safety evaluations.
The team also reviewed a sample of twenty 10 CFR 50.59 screenings for which Exelon had concluded that no safety evaluation was required.
These reviews were performed to assess whether Exelon's threshold for performing safety evaluations was consistent with 10 CFR 50.59. The sample included design changes, calculations, and procedure changes.The team reviewed the safety evaluations that Exelon had performed and approved during the time period covered by this inspection (i.e., since the last modifications inspection)not previously reviewed by NRC inspectors.
The 10 CFR 50.59 screenings were selected based on the safety significance, risk significance, and complexity of the change to the facility.In addition, the team compared Exelon's administrative procedures used to controlthe screening, preparation, review, and approval of safety evaluations to the guidance in NEI 96-07 to determine whether those procedures adequately implemented the requirements of 10 CFR 50.59. The safety evaluations and screenings reviewed by the team are listed in the Attachment.
b. Findinos No findings were identified.
.2.2.1 a..2.2 2 Permanent
Plant Modifications (12 samples)'A' Control Room Heatins, Ventilation and Air Conditioninq Svstem Modifications Inspection Scope The team reviewed modification 09-00708 that redesigned and reconfigured the turbine building closed cooling water (TBCCW) inlet and outlet piping to provide cooling water to the new'A' control room heating, ventilation and air conditioning (HVAC) condensers.
The new TBCCW piping was necessitated by an HVAC condenser upgrade (from coil coolers to Utube heat exchangers).
The previous TBCCW connections were welded directly to the coolers, whereas the new coolers have inlet and outlet connections that allow the use of threaded piping rather than welded connections.
The team reviewed the modification to verify that the design and licensing bases and performance capability of the 'A' control room HVAC system had not been degraded by the modification.
The team interviewed engineering staff and reviewed technical evaluations associated with the modification to determine if the 'A' control room HVAC system would function in accordance with the design assumptions.
The team performed several walkdowns of the 'A' control room HVAC system to independently assess Exelon's configuration control, TBCCW piping fit-up and supports, and the material condition of the HVAC components.
The team reviewed the associated post-modification test (PMT) results and recent'A'control room HVAC surveillance test results to verify that the system functioned as designed following the modification.
In addition, the team observed portions of an 'A' control room HVAC surveillance on February 10,2012, to verify the leak tightness of the TBCCW piping connections and the integrity of the ventilation boundary with the system in service. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1 R17.1 of this report. The team also reviewed corrective action issue reports (lRs) to determine if there were reliability or performance issues that may have resulted from the modification.
The documents reviewed are listed in the Attachment.
Findinqs No findings were identified.
Emerqencv Service Water Pipino Spool Replacements Inspection Scope The team reviewed modification 08-01040 that installed various piping spool pieces in the emergency service water (ESW) system in the reactor and turbine buildings.
Exelon targeted certain ESW piping spools for replacement based on identified internal degradation due to erosion and/or corrosion.
Based on pre-modification walkdowns, engineering determined that a one for one spool replacement was not practical due to existing obstructions as other piping and components were installed around the ESW piping over the years since original construction.
Engineering determined that shorter Enclosure a.
.2.3 3 length spool pieces would be needed for the modification, requiring
additional pipe flanges, resulting in increased loads on existing pipe supports and increased pipe stresses.The team reviewed the modification to verify that the design and licensing bases and structural integrity of the ESW piping and supports had not been degraded by the modification.
The team interviewed design engineers, and reviewed evaluations, pipe stress calculations, surveillance and PMT results, and associated maintenance work orders to verify that the ESW piping spool replacements were appropriately implemented and that the ESW piping was maintained in accordance with design assumptions.
The team also performed several walkdowns of the accessible portions of the modification to ensure that the system configuration was in accordance with design instructions and that ESW piping integrity was maintained.
The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1R17
.1 of this report. The team also reviewed corrective
action lRs to determine if there were reliability or performance issues that may have resulted from the modification.
The documents reviewed are listed in the Attachment.
Findinqs No findings were identified.
lntake Trash Rake Upqrade Modification Inspection Scope The team reviewed modification 06-00819 that replaced the existing rail mounted and conveyor debris removal system at the intake structure with a new upgraded trash raking system. The new system consists of bar racks, raking head and trolley, overhead monorail and associated support columns, dedicated trash dumpster, and an automated control system with remote emergency stop capability.
The previous rail mounted system proved difficult to operate and ineffective for collection and disposal of intake debris especially during periods of high debris accumulation.
Exelon upgraded the intake debris removal system with the newer, automated design to improve the system's effectiveness and reliability.
The team reviewed the modification to verify that the design and licensing bases and performance capability of the intake structure and its supported systems had not been degraded by the modification.
Specifically, the team reviewed calculations, technical evaluations, and operating procedures to verify that the overhead monorail and associated support columns would not adversely impact important to safety structures, systems, and components (SSC) at the intake during normal operation or under design basis conditions.
The team reviewed the associated work order instructions and documentation to verify that maintenance personnel implemented the modification as designed.
The team reviewed the associated PMT results, interviewed plant operators, and directly observed debris removal activities at the intake to verify proper operation of the upgraded system. The team also performed several walkdowns of the upgraded Enclosure 4 rake system and intake area SSCs to ensure that maintenance personnel installed the modification as designed, and to independently assess Exelon's configuration control and the material condition of the intake area. In addition, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1 R17.1 of this report. The team also reviewed corrective action lRs to determine if there were reliability or performance issues that may have resulted from the modification.
The documents reviewed are listed in the Attachment.
b. Findinqs No findings were identified.
.2.4 Emeroencv
Service Water Pump 'C' Discharoe Pipinq Modification a. lnspection Scope The team reviewed modification 11-00035 that installed an uncoated flanged tee downstream of the 'C' ESW pump discharge.
The piping tee was physically located beneath the intake structure deck. The piping tee replacement was emergent work necessitated due to an Exelon-identified through-wall leak in the existing piping tee.Exelon performed the replacement under a TS Limiting Condition for Operation that did not allow sufficient time for Exelon to obtain an internally coated tee. Accordingly, engineering evaluated the acceptability of using the uncoated piping tee until Exelon's planned replacement of all the ESW piping under the intake structure deck during the F all 2012 refueling outage.The team reviewed the modification to verify that the design and licensing bases and structural integrity of the ESW piping had not been degraded by the modification.
The team interviewed design engineers, and reviewed evaluations, non-destructive examination results, surveillance and PMT results, and associated maintenance work orders to verify that the ESW piping tee replacement was appropriately implemented, and that the ESW piping configuration supported continued operability through December 2012. The team also performed several walkdowns of the accessible portions of the modification to ensure that the system configuration was in accordance with design instructions and that ESW piping integrity was maintained.
The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1R17.1 of this report. The team also reviewed corrective action lRs to determine if there were reliability or performance issues that may have resulted from the modification.
The documents reviewed are listed in the Attachment.
b. Findinos No findings were identified.
b.a.2.6 5 Weak Link Analvsis for Ovster Creek Motor-Operated Valves Inspection Scope The team reviewed a modification 09-00889 that revised a weak link analysis (C-1302-900-E540-020)for several motor-operated valves (MOVs), including core spray system MOVs. The weak link analysis was revised to address valve stem material changes as well as a conversion to valve stems containing integral gages for MOV diagnostic testing for several of the MOVs.The team reviewed the weak link analysis to verify that the design and licensing bases and performance capability of the MOVs not been degraded by the change. The team interviewed engineering staff and reviewed the revised analysis to confirm the impacted systems would function in accordance with the design assumptions.
The team also reviewed the corrective action lR database to determine if there were reliability or performance issues that may have resulted from the modification.
ln addition, the team reviewed the associated equivalency change documentation that demonstrated that a formal 10 CFR 50.59 screen was not required.
The documents reviewed are listed in the Attachment.
Findinqs No findings were identified.
Service Water Cross-Connect from Reactor Buildinq Closed Coolinq Water Heat Exchanqer to Emerqencv Service Water Inspection Scope The team reviewed a modification 09-00433 that cross-connected the service water (SW) system reactor building closed cooling water (RBCCW heat exchanger discharge with the existing ESWSW cross-connect piping. The intent of the modification was to allow the replacement of the SW system piping downstream of the RBCCW heat exchanger leading into the SW system discharge header and prevent entry into high risk plant configuration during the refueling outage. This configuration allowed sufficient cooling to the RBCCW system while the SW normaldischarge piping was replaced.The team reviewed the modification to verify that the design and licensing bases and structural integrity of the associated piping had not been degraded by the modification.
The team interviewed design engineers, and reviewed evaluations, examination results, and associated completed maintenance activities to verify that the modified piping configuration supported continued functionality.
The team performed walkdowns of the modification to ensure that the system configuration was in accordance with design instructions.
The team also reviewed corrective action lRs to determine if there were reliability or performance issues that may have resulted from the modification.
In Enclosure 6 addition, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findinqs No findings were identified.
.2.7 Hardened Vent Valve Open Position to Permit Ventinq
a. Inspection Scope
The team reviewed modification 08-00864, which involved a calculation revision, to determine the smallest acceptable valve opening for the hardened vent system valves that would allow the required venting for both the drywell and torus. The stroke for the subject butterfly valves, V-23-13, -14, -15, and -16, had been limited to 75 degrees to allow them to close in the required stroke time, and which also considered valve structural limitations.
However, there was no tolerance associated with setting up the valve stroke. This calculation determined the minimum valve position, to be used for setup tolerance, to ensure that the vent capability would be satisfied.
The valves serve two purposes:
1) to close to provide containment isolation, and 2) to provide a hardened vent function.
The valves are now required to have a 70 degrees - 75 degrees band for the open position limit.The team reviewed the calculation to verify that the design and licensing bases and performance capability of the containment isolation and hardened vent functions of the subject valves had not been degraded by the modification.
Specifically, the team verified that design specifications remained valid for postulated scenarios.
The team interviewed engineers, and reviewed evaluations and completed surveillance and in-service test results to verify that the open position tolerances were appropriately implemented.
The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1R17.1 of this report. Finally, the team walked down the isolation valves with the system engineer to assess the material condition of the valves. The documents reviewed are listed in the Attachment.
b. Findinqs No findings were identified.
.2.8 Calculation
on Combustion Turbine Tank Oil Level a. lnspection Scope The team reviewed modification 10-00175 that revised calculation C-1302-743-E310-006 related to the required minimum fuel oil level for the Forked River Combustion Turbine (FRCT) fuel tank. The fuel reserve is required by contract with an outside organization to assure that the FRCTs can provide the power to Oyster Creek in the Enclosure 7 event of a station blackout.
The existing minimum level in the fueltank was 14 feet, but the revision changed the minimum level to 8 feet.The team reviewed the calculation to verify that the design and licensing bases and performance capability of the FRCT functions had not been degraded by the modification.
Specifically, the team verified that design specifications remained valid for the postulated station blackout scenario.
The team interviewed design engineers and reviewed the existing contract to ensure scenario assumptions remained valid. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1R17.1 of this report. Finally, the team walked down the FRCTS, including the fueltank and associated fuel supply system with the responsible engineer to assess the material condition of the FRCT system. The documents reviewed are listed in the Attachment.
b. Findinqs No findings were identified.
.2.9 Emerqencv
Diesel Generator Batterv Voltaqe for Control Circuits a. lnspection Scope The team reviewed modification 07-00552, which created a calculation to determine the minimum voltage available for both emergency diesel generator (EDG) breaker closing coils and associated control devices. The EDGs provide vital power to emergency buses during a loss-of-offsite power. Each EDG has a dedicated 120Vdc battery to provide starting power. The modification was performed because Exelon received a finding (NCV 05000219/2007006-01)in 2007 during a Component Design Basis Inspection performed by the NRC at Oyster Creek. The finding was related to an existing EDG battery sizing calculation that did not address the available DC voltage to the EDG breaker closing coils or associated control devices. The modification did not include any physical plant changes to the facility.The review was performed to verify that the design and licensing bases of the facility had not been degraded by the results of the new calculation.
The team reviewed the calculation and technical evaluations to assess whether the modification was consistent with design assumptions.
Power requirements were reviewed to verify that the EDG breaker closing coils and associated control circuits met the manufacturer's specifications.
Supporting electrical calculations and analyses for the EDG battery sizing requirements were reviewed to ensure design limits were not exceeded.
The team performed a walkdown of the EDG battery compartments to identify any abnormal conditions while in service. The team also conducted interviews with engineering staff to determine if the affected SSCs would function in accordance with the design assumptions.
Finally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the Attachment.
8 b. Findinqs No findings were identified.
.2.10 Revise Ovster Creek Short Circuit Studv
a. Inspection Scope
The team reviewed modification 07-00744 that updated the Oyster Creek Short Circuit Study (C-1302-700-5350-012).
The modification included converting the existing Short Circuit Study modeled in electricaltransient analysis software (DAPPER) to new electrical transient analysis software (ETAP). The new electrical distribution model also included corrections to reflect the as-built plant configuration.
An additional scenario, where the main generator provides power to the auxiliary transformer with one of the EDGs connected in parallelto an emergency bus for EDG testing, was also included.The modification did not include any physical plant changes to the facility.The review was performed to verify that the design bases and licensing bases of the facility had not been degraded by the short circuit study results. The results of the revised short circuit study showed that the interrupting and momentary fault currents were within the circuit breaker ratings of allthe 4160V Switchgear, 480V unit substations, and 480V motor control centers. There was also an improvement in calculated margin for the worst case scenario of a small break loss-of-coolant accident.Design assumptions were reviewed to evaluate whether they were technically appropriate and consistent with the UFSAR. The team also conducted interviews with engineering staff to determine if the affected SSCs would function in accordance with the design assumptions.
Finally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findinss No findings were identified.
.2.11 Plant Process Computer lntelliqent
Remote Control Unit Replacement
a. Inspection Scope
The team reviewed modification 08-00527 that upgraded the inpuUoutput components of the plant process computer.
Two intelligent remote control units (IRCU) and six expansion chassis were replaced with eight universal chassis. The existing inpuUoutput cards were transferred from the existing IRCU and expansion chassis to the new universal chassis. The modification was performed because the existing system used a custom interface that on rare occasion locked up the system and changed the scanned values to static values. The universal chassis interface with data from the following systems: Rod Worth Minimizer; 3D Monicore; Safety Parameter Display System;Enclosure I Emergency Response Data System; and Radioactive Gaseous Effluent Monitoring System.The review was performed to verify that the design and licensing bases and performance capability of the installed universal chassis had not been degraded by the modification.
Power requirements were reviewed to verify that the installed universal chassis met the manufacturer's specifications.
Replacement components and materials were reviewed to ensure that the modification conformed to the design specifications.
The team also verified that selected drawings and calculations were properly updated based on the installed universal chassis. The team reviewed the PMT to verify proper operation of the installed universal chassis. The team reviewed lRs associated with the universal chassis to verify that deficiencies were appropriately identified and corrected.
The team also conducted interviews with engineering staff to verify that the affected SSCs functioned in accordance with the design assumptions, and to verify the modification corrected the previously identified problem. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in section 1R17
.1 of this report. The documents
reviewed are listed in the Attachment.
b. Findinqs No findings were identified.
.2.12 Replace Recorder UR-622-24B
for Post-Accident Monitorinq Reactor Pressure and Level
a. Inspection Scope
The team reviewed modification 09-00590 that replaced an existing Yokogawa analog strip chart recorder with a new Yokogawa digital paperless recorder.
The new digital recorder displays the same plant inputs as the existing analog recorder.
The following data inputs are displayed on the new recorder:
Reactor Pressure; Reactor Level-Yarway; and Reactor Level- Fuel Zone. The modification was performed because the existing analog recorder was obsolete and spare parts were difficult to obtain. The modification included installation of the new recorder in the same location as the existing recorder and no additionalwiring was required.The review was performed to verify that the design and licensing bases and performance capability of the new digital recorder had not been degraded by the modification.
The team reviewed calculations and technical evaluations to assess whether the modification was consistent with design assumptions.
Power requirements were reviewed to verify that the new digital recorder met the manufacturer's specifications.
The replacement component was reviewed to verify that it was seismically qualified.
The team also verified that selected drawings, calculations, instrument calibration sheets, and procedures were properly updated based on the installation of the digital recorder.
The team reviewed the PMT to verify proper operation of the digital recorder.
The team reviewed lRs associated with the digital recorder to verify that deficiencies were appropriately identified and corrected.
In addition, the team Enclosure 4.10 reviewed the associated equivalency change documentation that demonstrated that a formal 10 CFR 50.59 screen was not required.
Finally, the team conducted interviews with engineering staff to determine if the affected SSCs would function in accordance with the design assumptions.
The documents reviewed are listed in the Attachment.
Findinss No findings were identified.
OTHER ACTIVITIES
ldentification and Resolution of Problems (lP 71152)Inspection Scope The team reviewed a sample of lRs associated with 10 CFR 50.59 and plant modification issues to determine whether Exelon was appropriately identifying, characterizing, and correcting problems associated with these areas, and whether the planned and/or completed corrective actions were appropriate.
In addition, the team reviewed lRs written on issues identified during the inspection to verify adequate problem identification and incorporation of the problem into the corrective action system.The lRs reviewed are listed in the Attachment.
Findinqs No findings were identified.
Meetinqs, includino Exit The team presented the inspection results to Mr. M. Massaro, Site Vice President, and other members of Exelon's staff at an exit meeting on February 10,2012. The team returned the proprietary information reviewed during the inspection and verified that this report does not contain proprietary information.
4c.42 40A6 b.Enclosure A-1 ATTACHMENT
=SUPPLEMENTAL
INFORMATION=
KEY POINTS OF CONTAGT Exelon Personnel
- A. Aganral, Design Engineer
- M. Benitez, Design Engineer
- J. Chrisley, Regulatory
Assurance
Specialist
- P. De, Design Engineer
- P. Desai, Design Engineer
- J. Flores, Design Engineer
- G. Malone, Director, Engineering
- M. Massaro, Site Vice-President
- K. Mayle, Design Engineer
- P. Procacci, Design Engineer
- H. Ray, Senior Manager, Design Engineering
- T. Ruggiero, System Manager
- S. Schwartz, System Manager
- J. Tabone, MOV/AOV Program Owner LIST OF ITEMS OPENED, CLOSED AND DISCUSSED None LIST OF DOCUMENTS
REVIEWED 10 CFR 50.59 Evaluations
OC-2009-E-0001, SW Cross-Connect
Downstream
of RBCCW Heat Exchangers, Rev. 1 OC-2010-E-0001, Application
of TRACG04P Version 4.2.60.3 for Stability
Analysis, Rev. 0 OC-2010-E-0002, Transition
to GNF-2 Fuel - lmpact on EAB, LPZ and CR Dose, Rev. 0 OC-2011-E-0001, MSO Bus 1D Appendix R Permissive
Switch lnstallation, Rev. 0 OC-2011-E-0002, Core Spray System MSO Modifications - MSO Scenarios
5k & 10b, Rev. 0 10 CFR 50.59 Screened-out
Evaluations
OC -2010-5-0046, Bank 5-6 Voltage Regulator
Controller
Modification, Rev. 0 OC -2010-S-0108, Component
Design Basis Inspection
Calculation
Revision, Rev. 0 OC -2011-5-0004, EDG-2 Breaker Logic Modification
for Appendix R Fire, Rev. 0 OC -2011-5-0016, RPS Sub-Channel
'1A'Alternate
MSIV Closure Trip Signal, Rev. 0 OC -2011-S-0017, lmplement
Level 2-3Data Diodes, Rev. 0 OC -2011-5-0048, Replacement
of EDG-1 Speed Switch, Rev. 0 OC -2011-S-0108, Core Spray Pump Trip Logic Mod - MSO Scenario 79, Rev. 0 Attachment
OC-2009-S-0093, 205.95.0 - Reactor Flood-up / Drain-down, Rev. 0 OC-2009-S-0170,310 - Containment
Spray System Operation, Rev. 0 OC-2009-S-0179,681.4.005 - Substation
Tour Sheet, Rev. 0 OC-2009-S-0181, ECR 09-00678 - Replace Recorders
LR-37/PR-53
and AR-1, Rev. 0 OC-2009-S-0184,316.1 - Condensate
Transfer System, Rev. 0 OC-2010-S-0026, 301.2 - Reactor Recirculation
System, Rev. 0 OC-2010-5-0027,654.2.002 - Installation
of Space Heater in Shutdown Cooling Room, Rev. 0 OC-2010-5-0048, ECR 10-00175 FRCT Fuel Oil System Evaluation
for SBO, Rev. 0 OC-2010-5-009355, ECR 10-00108 - TBCCW Heat Exchanger
Vent Bypass Line, Rev. 0 OC-2010-S-0190, Surveillance
665.5.00 Allow use of Hardened Vent for Venting, Rev. 1 OC-2011-5-0060, ABN-31 High Winds, Rev. 0 OC-2011-5-0070, ABN-32 Abnormal Intake Level, Rev. 0 OC-201 1-S-01 14. V-14-33 Steam Leak Condenser
ECR 1 1-00596. Rev. 0 Modification
Packaqes 06-00819, lntake Trash Rake Upgrade Modification, Rev. 0 07-00552, EDG Battery Voltage for Control Circuits, Rev. 0 07-00744, Revise Oyster Creek Short Circuit Study, Rev. 0 Q8-00527, Plant Process Computer Intelligent
Remote Control Unit Replacement, Rev. 2 08-00864, Hardened Vent Valve Open Position to Permit Venting, Rev. 0 08-01040, ESW System Piping Spool Replacements, Rev. 0 09-00433, SW Cross-Connect
from RBCCW Heat Exchanger
to ESW, Rev. 3 09-00590, Replace Recorder UR-622-24B
for Post-Accident
Monitoring
Reactor Pressure and Level, Rev. 0 09-00708, 'A'Control
Room HVAC System Modifications, Rev. 0 09-00889, Weak Link Analysis for'Oyster
Creek MOVs, Rev. 0 10-00175, Calculation
on Combustion
Turbine Tank Oil Level, Rev. 0 11-00035, ESW Pump 'C' Discharge
Piping Modification, Rev. 0 Calculations.
Analvsis.
and Evaluations
1164020-07, ESW Piping Leak Under Deck at lntake Equipment (Apparent
Cause), Rev. 0 1209774-02, ESW Pipe ES0106 lnspection
Technical
Evaluation, dated 6115111 1209774-04, ESW Pipe ES0112 Inspection
Technical
Evaluation, dated 6115111 1209774-05, ESW Pipe ES0212 Inspection
Technical
Evaluation, dated 6115111 943876-03, Intake and Dilution Trash Rake Short Hose Failures (Apparent
Cause), Rev. 0 42204999-05, CR HVAC'A'Condenser
Support Bracket Seismic Evaluation, dated 9l24l0g 42204999-15, Provide Guidance/Acceptance
forAs-Built
Condenser
Clamps, dated 11112109 42287516,'B'
lC Steam Inlet lsolation
Valve V-14-33 Packing Leakage Monitoring
Plan, Rev. 2 42296464-01, Leakage from 'A' Control Room HVAC Technical
Evaluation, dated 1131112 C-1302-168-E310-001, lntake Structure
Trash Rake Foundation
and Anchorages, Rev. 1 C-1302-251-E310-037, Pipe Stress Analysis, Fuel Pool Cooling Line NN-1, Model 2,from Skimmer Surge Tank to Pump Suction Nozzles NN01A/B and NN01C/D, Rev. 0 C-1302-532-E540-037, Piping Analysis - ESW System from Containment
Spray HX 3 & 4 through TB and Out to Yard, Rev. 1a C-1302-532-E540-041, Piping Analysis - ESW System from Containment
Spray HX1 &2 through RB to TB Entrance, Rev. 2a Attachment
C-1302-532-E540-045, Piping Analysis - ESW System from Containment
Spray HX 1 & 2 Outlets to Yard, Rev. 2a C-1302-700-5350-004, Oyster Creek Electrical
Model, Rev. 3 C-1302-700-5350-012, Oyster Creek Short Circuit Study, Rev. 3 C-1302-741-5350-009, Oyster Creek EDG Sizing Calculation, Rev. 1 C-1302-743-E310-006, Forked River Fuel Oil Transfer System Performance
Evaluation, Rev. 0 C-1302-822-5360-036, Hardened Vent System Capability, Rev. 2 C-1302-822-E310-083, EAB, LPZ, Control Room Doses due to LOCA, Rev. 0 C-1302-822-E310-084, EAB, LPZ, CR Doses due to Main Steam Line Break Accident, Rev. 0 C-1302-822-E310-085, EAB, LPZ, CR Doses due to Fuel Handling Accident, Rev. 0 C-1302-822-E310-086, EAB, LPZ, CR Doses due to Control Rod Drop Accident, Rev. 0 C-1302-900-E540-020, Weak Link Analysis Calculations
for Oyster Creek MOVs, Rev. 1A DRA CSW00661, Structural
Analysis for lntake Structure
Trash Rake Monorail System, Rev. 1 ECR 09-00685,204-45123
and 204-45611
'1O-Ton and 1S-Ton Gondensers, Rev.0 EXLNOC094-PR-01, Assessment
of the Oyster Creek Hardened Vent System, Rev. 0 lssue Reports 330592 350627 625029 6251 38 631025 645374 929813 943876 943993 945376 945676 953452 964555 972528 976547 978288 985986 991 345 1056623 1 058926 1059021 1062005 1 062500 1 088735 1115400 1 128036 1128042 1 135900 1164020 1 166208 1166220 1 1 66848 1 1 98623 1 198629 1243623 1243635 1276569 1 280593 1290865 1317861 1 31 8090" 1 31 8266*1 31 8288" 1319776*1319787*1321771*1 321 899" 1 321 996" 1322673 1323920*1323992*1324164 1324254" 1324795*1 31 8465*1323820.1324888*(* denotes NRC identified
during this inspection)
Drawinqs 13432.33-EM-1, Radiation
Shielding
Support Reactor Cavity Drain Line Pipe Supports, Rev. 10 13983-0002-E-01, Plant Process Computer System Network Block Diagram, Sh. 1 & 4, Rev. 0 13983-0002-E-07, Multiplexor
Circuit PC6 Subnet 'B' Switch Wiring Diagram, Rev. 0 15595.00-EM-1, Intake Structure
Modifications, Rev. 1 2153, Fuel Pool Cooling Filtering
& Drain Piping Plans and Sections Reactor Building, Rev. 4 2167, HVAC Control, Mechanical
Equipment
& Cable Room, Rev. 5 237E756, Spent Fuel Pool Cooling Flow Diagram, Rev. 53 3C-532-A3-1000, Pipe lntegrity
Inspection
Program ESW System Piping, Rev. 1 3D-531-22-1009, ECR 08-01040 Attachment
2, Sh. 1 , 2, & 3, Rev. 0 3D-532-24-001, Emergency
Service Water System Pipe Restraint
Modification, Rev. 0 3E-168-02-001 , General Arrangement
lntake Structure, Rev. 10 4031, Intake Structure
Sections and Details, Rev. 3 4034,Intake
Structure
Trash Rack and Stop Log Details, Rev. 3 538361, General Erection lntake Structure, Rev. G Attachment
557744, General Erection Intake Structure, Rev. D BR 2005, Reactor and Turbine Building Service Water System, Sh. 2, Rev. 105 BR 2006, Turbine Building Closed Cooling Water System Flow Diagram, Sh. 5, Rev. 58 BR 2010, Control and Cable Spreading
Rooms HVAC Flow Diagram, Rev. 32 BR 201 1, Reactor Building Ventilation, Sh. 2, Rev. 62 BR 30018,4160V
System One Line Diagram, Rev. 16 BR 3001C, 4160V System One Line Diagram, Rev. 1 BR 3005, Misc. Building 460V MCC One Line Diagram, Sh. 5, Rev. 10 GE 15786350, 480V System Electrical
Elementary
Diagram, Sh. 41, Rev. 23 GE 237E566, Reactor Protection
System, Sh. 17, Rev. 3 GU 3E-243-21-1000, Drywell and Torus Vacuum Relief System, Rev. 28 SN 13432.19-1, Nitrogen Supply System, Sh. 1, Rev. 33 U949-C-5000, Intake Structure
Plan, Rev. 1 Procedures
205.95.0, Reactor Flood-up / Drain-down, Rev. 18 2400-GMM-3900.52, Inspection
and Torquing of Bolted Connections, Rev. 5 307, lsolation
Condenser
System, Rev. 116 309.1, Turbine Building Closed Cooling Water System, Rev. 57 310, Containment
Spray System Operation, Rev. 98 312.11, Nitrogen System and Containment
Atmosphere
Control, Rev. 40 & 41 312.9, Primary Containment
Control, Rev. 52 & 53 344.2,lntake
Trash Rake Operation, Rev. 10 654.3.004, Control Room HVAC System 'A' Flow and Differential
Pressure Test, Rev. 13 665.5.001, Torus to Drywell Vacuum Relief Valve Leak Rate Test, Rev. 29 681.4.005, Substation
Tour Sheet, Rev. 17 & 18 ABN-31, High Winds, Rev. 14 & 16 ABN-32, Abnormal lntake Level, Rev. 18 AD-AA-01, Processing
of Procedures
and T&RMs, Rev. 23 CC-AA-1, Configuration
Control, Rev. 0 CC-AA-10, Configuration
Control Process Description, Rev. 6 CC-AA-102, Design lnput and Configuration
Change lmpact Screening, Rev.22 CC-AA-103, Configuration
Change Control for Permanent
Physical Plant Changes, Rev.22 CC-AA-107, Configuration
Change Acceptance
Testing Criteria, Rev. 8 CC-AA-107-1001 , Post Modification
Acceptance
Testing, Rev. 4 CC-AA-13, Margin Management, Rev. 2 CC-MA-102-1001, Design Inputs and lmpact Screening - lmplementation, Rev. 9 LS-AA-104, Exelon 50.59 Review Process, Rev. 6 LS-AA-104-1000, Exelon 50.59 Resource Manual, Rev. 6 LS-AA-120, lssue ldentification
and Screening
Process, Rev. 14 MA-AA-743-310, Diagnostic
Testing and Evaluation
of Air Operated Valves, Rev. 5 NRT-OC-08-0006, RTP-2000 Functional
Test, Rev. 0 OP-OC-108-109-1001, Severe Weather Preparation
T&RM for Oyster Creek, Rev. 12 Attachment
Work Orders A2149453 p.2204999 A2262068 42262069 A2270528 A2270529 c2014505 c2014848 c2016918 c2017315 c2019099 c2021995 c2023483 c2025008 c2025252 c2025388 c2025389 M2119029 R0802188 R2127181 R2128890 R2132325 R2156772 R2162526 R2165640 R2185877 R2189853 Miscellaneous
AWC Flat Festoon Cable (PVC) Specifications, dated 1125112 C2017315, Post-lnstallation
Walkdown Checklist, dated 1 1 19110 C2019099-18, SQUG Walkdown - Seismic Adequacy per ECR 09-00708, dated 11l13l}g CR HVAC System Walkdown Report, dated 8130111 & 11129111 Cycle 23 Core Operating
Limits Report - Oyster Creek, Rev. 5 ECR OC 06-00819 Attachment
F, Acceptance
Test Criteria for Intake Structure
Trash Raking System, Rev. 0 First Amendment
to Station Blackout Agreement
between Forked River Power LLC and Exelon Generation
Company, LLC, dated 5112110 GE-NE-0000-0052-5690-R0, TRACGO4 10 CFR 50.59 Evaluation
Basis, Rev. 0 GS 04L43B01-01E, Daqstation
DX1000N General Specifications, Rev.5 HVAC - Air Handling Equipment
PCM Template, dated 815111 f n-service
Testing Bases Document (V-23-13), January 2012 NEDC-33065P, Application
of Stability
Long Term Solution Option 2 to Oyster Creek, Rev. 0 OCNGS - Relief from the Requirements
of the ASME Code, Relief Request No. VR-02 for the Fifth Inservice
Testing Interval (TAC No. ME7618), dated 1124112 OCNGS - Relief Request RP-04, Regarding
SW Pump Suction Pressure Gages, and RV-51, Containment
lsolation
Valve Position Indication (TAC No. M84945), dated 1012102 Program Health Report, NRC Generic Letter 89-13, Q4-2011 Submittal
of Proposed Alternative
and Relief to the Requirements
of 10 CFR 50.55a Concerning
the Fourth Ten-Year Interval In-service
Testing Program, dated 4119102 Submittal
of Relief Request for the Fifth ln-service
Testing lnterval (RA-11-089), dated 11117111 System Health Report, Circulating
Water, Q3-2011 System Health Report, Control Room HVAC, Q4-2011 System Health Report, Emergency
Service Water, Q4-2011 System Health Report, Screen Wash, Q3-2011 System Health Report, Service Water, Q3-2011 V-14-33 Motor Temperature
Trend Data, dated 10130111 - 1122112 V-14-33 Packing Leak Rate Trend Data, dated 10110111 - 1116112 VM-OC-0008, Magne-Blast
Circuit Breaker Vendor Manual, Rev. 14 VM-OC-2888, lntake Trash Rake Installation
Operation
and Maintenance
Manual, Rev. 1 Design & Licensinq
Bases NRC Regulatory
Guide 1.183, Alternate
Radiological
Source Terms for Evaluating
Design Basis Accidents
at Nuclear Power Reactors, July 2000 OCNGS Updated Final Safety Analysis Report, Rev. 17 Attachment
Safety Evaluation
by the Office of Nuclear Reactor Regulation
Related to Amendment
No. 262to Facitity Operating
License No. NPF-16 for the Oyster Creek Nuclear Generating
Station -Application
of Alternate
Source Term Methodology (TAC No. MC6519), dated 4126107 SDBD-OC-243, Design Basis Document for Containment
System, Rev. 1 TDR No. 1099, Station Blackout Evaluation
Report, Rev. 4 NUREG-0800
SRP-15.0.1, Radiological
Consequence
Analyses Using Alternate
Source Terms Standard Review Plan, Rev. 0 PBD-AMP-3.2.04, Oyster Creek License Renewal Project Periodic Inspection
of Ventilation
Systems, Rev. 1 TAC No. 68577, Safety Evaluation - SBO Analysis OCNGS, dated 8123191 Completed
Surveillance
and Modification
Acceptance
Tests 2011-002-001, ESW 2 Piping, Downstream
of P-33C UT NDE Report, performed
21111 2011-002-015, ESW Piping Under lntake (ES0106) UT NDE Data Report, performed
27111 2011-002-016, ESW Piping Under Intake (ES0112) UT NDE Data Report, performed
28111 2011-002-017, ESW Piping Under Intake (ES0212) UT NDE Data Report, performed
29111 2400-SMM-3900.04
Exhibit 1, C2017315 ESW 2 Pressure Test (ASME Xl), performed
1117110 2400-SMM-3900.04
Exhibit 1, C2025008 ESW Tee Pressure Test (ASME Xl), performed
24111 2400-SMM-3900.08
Exhibit 1, C2023483-13
General Hydrostatic
Test, Initial Service Leak Test, and Pneumatic
Test (ANSI 831.1), performed
11130110 607.4.017, CS/ESW Pump System 2 Operability
and Quarterly
lnservice
Test, performed
20112 636.2.012, Diesel Generator#1
Battery Service Test, performed3lTlll
636.2.013, Diesel Generator
- 2 Battery Service Test, performed
10/9/09 & 12113110 636.4.001, Diesel Generator
- 1 Automatic
Actuation
Test, performed
11124110 636.4.002, Diesel Generator
- 2 Automatic
Actuation
Test, performed
11112110 654.3.004, Control Room HVAC 'A' Flow and DP Test, performed
6/28/07,6123109, &6128111 654.3.006, Control Room HVAC 'B' Flow and DP Test, performed
6117110 654.4.003, Control Room HVAC System Operability
Test, performed
1112112 665.5.001, Torus to Drywell Vacuum Relief Valve Leak Rate Test, performed
1129110 678.4.001, Primary Containment
lsolation
Valve Operability
and lST, (V-23-13, -14, -15, -16), performed
4113111, 812111, & 10112111 681.4.005, Substation
Tour Sheet, performed
25112 - 1127112 C2019099-10, In-service
Leak Test for 3 New Condensers
Replaced by ECR 09-00708, performed
11l13l0g C2021995-07, ln-Service
Leak Test for Condenser
Refrigerant
Pipe Assembly, Compressor, Evaporator
and allAssociated
Air Lines, performed
11116109 ER-AA-335-015
1, C2025008 l/'I-2, NDE Report, performed
24111 Weld Map No. 532-WM-050/0, C2017315 Pipe Weld Record Sheet, dated 11124110 Weld Map No. 532-WM-060/0, C2017315 Pipe Weld Record Sheet, dated 1211110 Attachment
ADAMS CFR DC DRS EDG Exelon ESW FRCT HVAC IP IR IRCU MOV NEI NRC OCNGS PARS PMT RBCCW SSC SW TBCCW TS UFSAR Vdc A-7 LIST OF ACRONYMS Agencywide
Documents
Access and Management
System Code of Federal Regulations
Direct Current Division of Reactor Safety Emergency
Diesel Generator Exelon Nuclear Northeast Emergency
Service Water Forked River Combustion
Turbine Heating, Ventilation
and Air Conditioning
lnspection
Procedure lssue Report Intelligent
Remote Control Unit Motor-Operated
Valve Nuclear Energy Institute Nuclear Regulatory
Commission
Oyster Creek Nuclear Generating
Station Publicly Available
Records Post-Modification
Test Reactor Building Closed Cooling Water Structure, System, and Component Service Water Turbine Building Closed Cooling Water Technical
Specifications
Updated Final Safety Analysis Report Volts, Direct Current Attachment