IR 05000219/2012007
| ML120790052 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/16/2012 |
| From: | Doerflein L Engineering Region 1 Branch 2 |
| To: | Pacilio M Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-12-007 | |
| Download: ML120790052 (22) | |
Text
UNITED STATES
ffi NUCLEAR REGULATORY COIUIUI ISSION
REGION I
475 ALLENDALE ROAD KING OF PRUSSIA, PENNSYLVANIA 19406-1415 l.larch 76, 2AL2 Mr. Michael 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 MAttachment: Supplemental Information
REGION I Docket No.: 50-219 License No.: DPR-16 Report No.: 0500021 9/2012007 Licensee: Exelon Nuclear Facility: Oyster Creek Generating Station Location: Forked River, New Jersey Inspection Period: January 23,2012 through February 10,2012 Inspectors: S. Pindale, Senior Reactor lnspector, Division of Reactor Safety (DRS),
Team Leader J. Schoppy, Senior Reactor Inspector, DRS J. Rady, Reactor lnspector, DRS Approved By: Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Enclosure
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 Permanent Plant Modifications (12 samples)
.2.1 'A' Control Room Heatins, Ventilation and Air Conditioninq Svstem Modifications
a. 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.
.2.2 Emerqencv Service Water Pipino Spool Replacements
a. 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 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.
.2.3 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 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.
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.
b. Findinqs No findings were identified.
2.6 Service Water Cross-Connect from Reactor Buildinq Closed Coolinq Water Heat Exchanqer to Emerqencv Service Water
a. 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 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 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.
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; 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 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
4c.42 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.
b.
Findinqs No findings were identified.
40A6 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.
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
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
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
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
209774-02, ESW Pipe ES0106 lnspection Technical Evaluation, dated 6115111
209774-04, ESW Pipe ES0112 Inspection Technical Evaluation, dated 6115111
209774-05, ESW Pipe ES0212 Inspection Technical Evaluation, dated 6115111
943876-03, Intake and Dilution Trash Rake Short Hose Failures (Apparent Cause), Rev. 0
204999-05, CR HVAC'A'Condenser Support Bracket Seismic Evaluation, dated 9l24l0g
204999-15, Provide Guidance/Acceptance forAs-Built Condenser Clamps, dated 11112109
287516,'B' lC Steam Inlet lsolation Valve V-14-33 Packing Leakage Monitoring Plan, Rev. 2
296464-01, 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
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 964555 1115400 1276569 1322673
350627 972528 1 128036 1 280593 1323920*
25029 976547 1128042 1290865 1323992*
251 38 978288 1 135900 1317861 1324164
631025 985986 1164020 1 31 8090" 1324254"
645374 991 345 1 166208 1 31 8266* 1324795*
29813 1056623 1166220 1 31 8288" 1 31 8465*
943876 1 058926 1 1 66848 1319776* 1323820.
943993 1059021 1 1 98623 1319787* 1324888*
945376 1062005 1 198629 1321771*
945676 1 062500 1243623 1 321 899"
953452 1 088735 1243635 1 321 996"
(* 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
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
2.11, Nitrogen System and Containment Atmosphere Control, Rev. 40 & 41
2.9, 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
Work Orders
A2149453 c2014848 c2025252 R2132325
p.2204999 c2016918 c2025388 R2156772
A2262068 c2017315 c2025389 R2162526
262069 c2019099 M2119029 R2165640
A2270528 c2021995 R0802188 R2185877
A2270529 c2023483 R2127181 R2189853
c2014505 c2025008 R2128890
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
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 1121111
2011-002-015, ESW Piping Under lntake (ES0106) UT NDE Data Report, performed 4127111
2011-002-016, ESW Piping Under Intake (ES0112) UT NDE Data Report, performed 4128111
2011-002-017, ESW Piping Under Intake (ES0212) UT NDE Data Report, performed 4129111
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 1124111
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 1120112
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 1 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 1125112 - 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 Attachment 1, C2025008 l/'I-2, NDE Report, performed 1124111
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
LIST OF ACRONYMS
ADAMS Agencywide Documents Access and Management System
CFR Code of Federal Regulations
DC Direct Current
DRS Division of Reactor Safety
EDG Emergency Diesel Generator
Exelon Exelon Nuclear Northeast
ESW Emergency Service Water
FRCT Forked River Combustion Turbine
HVAC Heating, Ventilation and Air Conditioning
IP lnspection Procedure
IR lssue Report
IRCU Intelligent Remote Control Unit
MOV Motor-Operated Valve
NEI Nuclear Energy Institute
NRC Nuclear Regulatory Commission
OCNGS Oyster Creek Nuclear Generating Station
PARS Publicly Available Records
PMT Post-Modification Test
RBCCW Reactor Building Closed Cooling Water
SSC Structure, System, and Component
TBCCW Turbine Building Closed Cooling Water
TS Technical Specifications
UFSAR Updated Final Safety Analysis Report
Vdc Volts, Direct Current
Attachment