IR 05000334/2010006

Download: ML102150492

Text

Mr. Paul Harden Site Vice President August 3, 2010 FirstEnergy Nuclear Operating Company Beaver Valley Power Station P.O. Box 4, Route 168 Shippingport, PA 15077

SUBJECT: BEAVER VALLEY POWER STATION -NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000334/2010006 AND 05000412/2010006

Dear Mr. Harden:

On June 24,2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Beaver Valley Power Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on June 24, 2010, with Mr. Ray Lieb, Director Site Operations, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license. In 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 of significance were identified. In 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 (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html(the Public Electronic Reading Room). Docket Nos: 50-334, 50-412 License Nos: DPR-66, NPF-73

Sincerely,Lawrence T. Doerflein, Chi Engineering Branch 2 Division of Reactor Safety

Enclosure:

Inspection Report No. 05000334/2010006; 05000412/2010006

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Supplemental Information cc w/encl: Distribution via ListServ ,

Mr. Paul Harden Site Vice President August 3, 2010 FirstEnergy Nuclear Operating Company Beaver Valley Power Station P.O. Box 4, Route 168 Shippingport, PA 15077

SUBJECT: BEAVER VALLEY POWER STATION -NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000334/2010006 AND 05000412/2010006

Dear Mr. Harden:

On June 24, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Beaver Valley Power Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on June 24, 2010, with Mr. Ray Lieb, Director Site Operations, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license. In 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 of significance were identified. In 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 (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html(the Public Electronic Reading Room). Docket Nos: 50-334, 50-412 License Nos: DPR-66, NPF-73

Sincerely,IRA! Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety

Enclosure:

Inspection Report No. 05000334/2010006; 05000412/2010006

w/Attachment:

Supplemental Information cc w/encl: Distribution via ListServ ADAMS ACCESSION NO. ML 102150492 SUNSI Review Complete: Initials) DOCUMENT NAME: Y:\Division\DRS\Engineering Branch 2\Arner\beaver valley mods\BVModsReport2010006.doc After declaring this document "An Official Agency Record" it will be released to the Public. To receive a copy of this document, Indicate In the box: "ell C = Opy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE RI/DRS I RI/DRP I RI/DRS I I NAME FArnerl RBellamvl LDoerfieinl DATE 07/23/10 07/30/10 08/3110 OFFICIAL RECORD COpy Distribution w/encl: M. Dapas, Acting RA (R10RAMAIL Resource) D. Lew, Acting DRA (R10RAMAIL Resource) J. Clifford, DRP (R1DRPMAIL Resource) D. Collins, DRP (R1DRPMAIL Resource) D. Roberts, DRS (R1DRSMaii Resource) P. Wilson, DRS (R1DRSMaii Resource) R. Bellamy, DRP S. Barber, DRP C. Newport, DRP D. Werkheiser, SRI E. Bonney, RI P. Garrett, Resident OA L. Trocine, RI OEDO D. Bearde, DRS RidsNRRPMBeaverValley Resource RidsNrrDorlLpl1-1 Resource@nrc.gov ROPreportsResource@nrc.gov L. Doerflein, DRS F. Arner, DRS U.S. NUCLEAR REGULATORY COMMISSION REGION I Docket No.: 50-334, 50-412 License No.: DPR-66, NPF-73 Report No.: 05000334/2010006 and 05000412/2010006 Licensee: FirstEnergy Nuclear Operating Company (FENOC) Facility: Beaver Valley Power Station, Units 1 and 2 Location: Post Office Box 4 Shippingport, PA 15077 Inspection Period: June 7 through June 24, 2010 Inspectors: F. Arner, Senior Reactor Inspector, Division of Reactor Safety (DRS), Team Leader L. Scholl, Senior Reactor Inspector, DRS E. Burkett, Reactor Inspector, DRS A. Dugandzic, Reactor Engineer, DRS (in-training) Approved By: Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000334/2010006; IR 05000412/2010006; 6/07/2010 -6/24/2010; FirstEnergy Nuclear Operating Company (FENOC); Beaver Valley Power Station, Units 1 & 2; 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 and one inspector in training. No findings of significance were identified. 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. NRG-Identified and Self-Revealing Findings No findings of significance were identified.

B. Licensee-Identified Violations

None. ii

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1 R 17 Evaluations of Changes. Tests. or Experiments and Permanent Plant Modifications (IP 71111.17)

.1 Evaluations of Changes. Tests. or Experiments (28 samples)

a. Inspection Scope

The team reviewed four 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 FENOC had been required to obtain NRC approval prior to implementing the change. The team interviewed plant staff and reviewed supporting information including calculations, analyses, design change documentation, procedures, the UFSAR, the Technical Specifications (TSs), 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 (NEI) 96-07, " Guidelines for 10 CFR 50.59 Evaluations," as endorsed by NRC Regulatory Guide 1.187, "Guidance for Implementation 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-four 10 CFR 50.59 screenings for which FENOC had concluded that no safety evaluation was required. These reviews were performed to assess whether FENOC'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 FENOC had performed and approved during the time period covered by this inspection (Le., since the last modifications inspection) not previously reviewed by NRC inspectors. The screenings were selected based on the safety significance, risk significance, and complexity of the change to the facility. In addition, the team compared FENOC's administrative procedures used to control the 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 reviewed safety evaluations and screenings are listed in the attachment.

b. Findings

No findings of significance were identified. Enclosure 2

.2 Permanent Plant Modifications (9 samples) .2.1 Unit 1 Emergency Diesel Generator (EDG) K 1 Relay Replacement*

a. Inspection Scope

The team reviewed a modification (Engineering Change Package (ECP) 08-0095) that replaced the original latching style relays with an electrically equivalent relay without the latching feature. The K1 relay design function is to shunt the field circuit during the EDG shutdown sequence and thereby terminate the generator electrical output in a controlled manner. The original relay would remain latched closed until unlatched during the next EDG startup sequence. One reason for the replacement was that the original style relays were found to be potentially subject to auxiliary contact terminals coming loose during a seismic event. Additionally, Unit 2 had experienced a problem with this latching type relay when it failed to release during an EDG start sequence and thereby rendered the EDG inoperable. This modification also added indicating lights to allow operators to verify the K1 relay opened as designed at the end of the shutdown sequence indicating that the excitation system was properly aligned for the next EDG start. An indicator light was also added to allow monitoring of the K2 field flash relay in a similar manner. The team conducted the review to verify that the design bases, licensing bases, and performance capability of the EDGs were not degraded by the component replacements and circuit modifications. The team discussed the modification and design basis with design and system engineers, and reviewed the scope and results of post-modification testing to assess the capability of the EDGs to perform their safety function during a design basis event. The team also confirmed that surveillance tests, operational procedures, and drawings had been appropriately updated to reflect the design change. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.2 Unit 1 Containment Sump Discharge Flow Transmitter Replacement

a. Inspection Scope

The team reviewed a modification (ECP-10-00010) that replaced an obsolete containment sump discharge flow instrumentation system with an equivalent model. The scope of the modification included the replacement of the flow tube and flow transmitter, and changed the location of its electrical power feed. The instrument provides a method of determining the amount of unidentified leakage from systems located in the containment building as required by plant technical specification 3.4.15, RCS Leakage Detection Instrumentation. Enclosure 3 The team conducted the review to ensure that the design bases, licensing bases, and performance capability of the leakage monitoring system had not been adversely affected by the modification. The team reviewed the associated work order packages, and conducted interviews with design and system engineers regarding the design, installation, calibration, and testing of the instrumentation to verify that the modification was adequate. The team walked down the accessible portions of the new equipment to ensure the system configuration was in accordance with the design instructions. The team also verified that drawings and operating procedures had been appropriately revised to reflect the change. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.3 Reclosing Reactor Trip Breaker Under Test

a. Inspection Scope

The team reviewed a modification (ECP-08-0134) performed by FENOC to install a local test switch in the reactor trip switchgear. The test switch allows local closing of the individual reactor trip breakers during surveillance testing instead of both trip breakers (the breaker under test and the other in-service breaker) receiving a closing signal when the breaker is closed from the control room switch. As a result, the in-service breaker is not unnecessarily subjected to the reverse inductive voltage spike during testing. The team evaluated the change to confirm that the design bases, licensing bases, and performance capability of the reactor trip system would not be affected by the change. The team also walked down accessible portions of the system to assess the configuration and material condition of the system, and reviewed updated surveillance procedures to ensure they included appropriate directions for the technicians operating the switches during testing. The team interviewed design engineers to review the design change and the potential impact on proper system operation. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified. Enclosure 4

.2.4 Refueling Water Storage Tank (RWST) Level Interlock to Recirculation Spray (RS) Pump Start

a. Inspection Scope

The team reviewed a modification (ECP 06-0227) that revised the automatic start signal for the RS pumps at both units to be initiated from an RWST level input. FENOC implemented the modification to regain RS pump net positive suction head (NPSH) margin that was reduced following the installation of modified containment sump strainers, which had been installed to address potential debris accumulation. Prior to this modification, the RS pumps started automatically following a containment high-high pressure signal and a time delay. The modification included adding main control room benchboard status lights and changing electrical wiring for the RS pump start circuit and relay logic. The team conducted the review to verify that the design bases, licensing bases and performance capability of the recirculation spray pumps had not been adversely affected by the modification. The team reviewed drawings, calculations, operator training modules, and procedures to ensure they had been properly updated to incorporate the changes to the RS pump start logic. In addition, post-modification testing was reviewed to verify the logic sequence and automatic RS pump start were appropriately tested. The team walked down portions of the modification to verify that status lights and annunciators were added to the main control room benchboards. The team also discussed the modification and design basis with design engineers to assess the adequacy of the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R 17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of Significance were identified . . 2.5 Recirculation Spray Heat Exchanger (HX) Inputs for Modular Accident Analysis Program (MAAP) Containment Analysis

a. Inspection Scope

The team reviewed an analysis (ECP 09-03277) performed by FENOC to evaluate the fouling factor of the Unit 2 RS heat exchangers and the potential for air entrapment within the heat exchangers. This analysis was performed to ensure the design fouling factor assumption for the two RS heat exchangers cooled by service water was appropriate. FENOC performed the analysis using the Electric Power Research Institute (EPRI) Modular Accident Analysis Program (MAAP) to demonstrate the fouling factor was acceptable for component performance during design basis accident conditions. Additionally, FENOC evaluated the internal shell side velocities of the heat exchangers to determine if they were sufficient to sweep air from the system during Enclosure 5 initial fill. Using a Froude number criterion, FENOC determined that air would be swept from the system upon initial fill during system initiation, and not impede HX performance. The team reviewed the analysis and associated calculations to verify the assumptions and parameters used were valid and considered worst case scenarios. The parameters included service water temperature, containment temperature and pressure, number of plugged tubes, sump temperature, and heat exchanger baffle arrangement. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.6 Turbine Driven Auxiliary Feedwater (TDAFW) Pump Steam Supply Solenoid Operated Valves (SOVs) -Benchboard Control Switch Replacements

a. Inspection Scope

The team reviewed a modification (ECP 07-0076) that replaced the selector switches for the TDAFW pump steam admission solenoid operated valves in the Unit 2 main control room. FENOC implemented this modification to address a postulated accident scenario such as a steam generator tube rupture (SGTR) event which would require plant operators to shut down the TDAFW pump from the control room using class 1 E safety related controls. Prior to the modification, the selector switches had" return-to-center" operators that prevented the steam admission valves from maintaining a closed position which may be necessary during certain design basis events. The replacement switches were designed to maintain the position in which they have been placed. The modification included installing an annunciator in the main control room, and replacing and rewiring the control switches. The team reviewed operating procedures, operator training modules, and drawings to ensure they had been properly updated. The team reviewed work order packages and conducted interviews with design and system engineers regarding the design, installation, and testing to verify that the modification was adequate. The team performed a walkdown of the modification in the main control room and discussed the alarm response procedure with the operators. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified. Enclosure 6

.2.7 Unit 2 Joint Owners Group (JOG) Motor Operated Valve (MOV) Periodic Verification Program Implementation

a. Inspection Scope

The team reviewed a modification (ECP 08-0504) that changed the setup for motor operated valve (MOV) actuators associated with the Unit 2 safety injection (SI) system. FENOC implemented this modification to incorporate requirements of the JOG MOV program which is an industry-wide response to Generic Letter (GL) 1996-05, " Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves." The team selected a sample of two valves to review the changes performed. The valves reviewed were 2SIS-MOV867 A; high head SI valve, and 2SIS-MOV869A, hot leg SI isolation valve. The actuator control switch setup was modified to account for increased torque and thrust requirements, and reduced stress relative to the weak link components. Additionally, the control switch for the closure function for valve 2SIS-MOV867 A was changed from torque switch control to limit switch control. The team conducted the review to verify that the design bases and performance capability of the safety injection system had not been adversely affected by the modification. The team reviewed associated drawings and calculations to ensure they were properly updated and the correct assumptions were used. In addition, the team reviewed completed work orders to verify the appropriateness of the testing and acceptance criteria. The team also discussed the modification and design bases with design and system engineers to assess the adequacy of the modification. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.8 Unit 1 Quench Spray Loop Seal Modification

a. Inspection Scope

The team reviewed a modification (ECP 07-0145) associated with the installation of a plug in the vent hole on the two quench spray (as) system loop seals and the repositioning of drain valves in the QS piping from open to normally closed. The modification was implemented in order to prevent air ingestion into the recirculation spray (RS) and low head safety injection (LHSI) pumps through the containment sump plenums. This was required due to the revised containment sump design, which had created the potential for air ingestion from the as system. Specifically, several of the as branch lines penetrate and discharge into the containment sump plenum. The design intent of those lines was to increase NPSH margin by adding the cooler QS water initially to the containment-sump. This modification also identified that the QS Enclosure 7 piping loop seal to the containment sump would be required to be kept full to prevent air ingestion during a postulated accident which would require containment sump recirculation. The team reviewed the associated calculations to ensure the methods used were reasonable in determining that the loop seal would be able to prevent air ingestion during postulated accident scenarios. Specifically, the team reviewed the design input assumption for containment pressure to ensure the calculated height of the loop seal would be adequate at the time the as pumps are secured after RWST depletion. The team reviewed emergency operating procedures to verify that the containment pressure design input was reasonable. Additionally, the team reviewed operating procedures established to ensure that the as loop seal would be adequately filled prior to operation as well as FENOC' s assessment for potential leakage impacting the design conclusions. The team also reviewed the response time change associated with the as system because of the change regarding the loop seal being maintained water solid prior to operation. This review was performed to ensure that surveillance tests and calculations had been revised appropriately. The team interviewed design engineers to review the design change and the potential impact on proper system operation. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified . . 2.9 Revision to Maximum Differential Pressure Across the Category I Motor Operated Valves in the Unit 2 Safety Injection System

a. Inspection Scope

The team reviewed a revision to design calculation 10080-N-584 that changed the differential pressure (DP) and line pressure requirements for SI valves due to the replacement of the Unit 2 charging pump rotating assemblies. The team reviewed FENOC' s methodology used in determining the revised DP parameters for various valves within the system. The team reviewed selected design inputs used in the revised calculation to verify they were reasonable. The design inputs verified included the static pressure affect from the refueling water storage tank, the containment pressure assumption at the time of swapover to sump recirculation, and the static elevation differences between the pumps and valves. The charging pump performance curve design input was also verified to have been correctly incorporated within the calculations with respect to the change of performance due to the replaced rotating assembly. The emergency operating procedures were reviewed to verify equipment alignments assumed in the determination of valve differential pressures were in accordance with procedure direction. Enclosure 8 The team sampled a few of the revised design DP outputs to ensure that the information was appropriately translated into torque calculations for Unit 2 SI valve 2SIS-MOV8811A, which automatically opens to provide a recirculation path from the containment sump via the recirculation pumps, during the recirculation phase of-safety injection. Additionally, the team verified that the revised DP output was appropriately translated to the torque calculation for 2SIS-MOV863B, which automatically opens during the recirculation phase to divert the flow of the recirculation spray pumps to the suction of the high head safety injection (HHSI)/charging pumps. The team performed a walkdown of the SI inlet isolation valve, to assess the material condition of the valve. The team also performed interviews with the responsible motor operated valve engineers to determine whether the design changes appeared reasonable. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

40A2 Identification and Resolution of Problems (IP 71152)

a. Inspection Scope

The team reviewed a sample of the condition reports (CRs) associated with 10 CFR 50.59 and plant modification issues to determine whether FENOC was appropriately identifying, characterizing, and correcting problems associated with these areas, and whether the planned or completed corrective actions were appropriate. In addition, the team reviewed CRs written on issues identified during the inspection to verify adequate problem identification and incorporation of the problem into the corrective action system. The CRs reviewed are listed in the attachment.

b. Findings

No findings of significance were identified. 40A6 Meetings, including Exit The team presented the inspection results to Mr. Ray Lieb, Director, Site Operations, and other members of FENOC's staff at an exit meeting on June 24, 2010. The team returned the proprietary information reviewed during the inspection to the licensee and verified that this report does not contain proprietary information. Enclosure A-1 ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

FENOC Personnel

R. Fedin, Regulatory Compliance

C. Mancuso, Design Engineering Manager

D. Price, Design Engineering Supervisor

K. Woessner, Nuclear Staff Engineer

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED None

LIST OF DOCUMENTS REVIEWED

10

CFR 50.59 Evaluations 07-00157, Impact of Reduced Atmospheric Relief Dump Valve (ASDV) Capacity, Rev. 0 07 -05204, Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project, Rev. 0 08-05643, BV2 Primary Zinc Addition, Rev. 0 09-02468, BV2 UFSAR Change -Containment Sump Screen Passive Failure, Rev. 0 10

CFR 50.59 Screened-out Evaluations 07-00275, Reactor Coolant Temperature Loop 2RCS-T412 Delta T-Tavg Protection Channell Calibration, Rev. 0 07-00521, Containment Coatings Walkdown, Rev. 0 07-05146, 18-Month Slave Relay Testing (Train B), Rev. 0 08-02593, Elimination of Recirculation Spray Pump Response Time SR 3.3.2.9, Rev. 1 08-05721, Replace Valves 2RCS-633 & 2RCS-634 with Pipe Spool Piece, Rev. 0 09-03164, Quench Spray System Trouble, Rev. 0 09-04690, RHS Fill and Vent from RWST, Rev. 0 08-05359, BV2 EDG ASP Transfer Circuit Relay Change, Rev. 0 09-00048, Station Battery Jumper Installation and Removal, Rev. 0 09-00238, Calculation 8700-DMC-1430 Revision, Rev. 0 09-00374, Incipient Fault Monitor Repair, Rev. 0 09-01089, 1

OST-1.22 Procedure Revision, Rev. 0 10-00126, Recirculation Spray Pump Test, Rev. 0 10-01483, Residual Heat Removal System Train A Valve Exercise, Rev. 0 10-00326, 1

MSP-9.05-1 Calibration Procedure Revision, Rev. 0 10-00342, Emergency Diesel Generator Monthly Test Procedure Change, Rev. 0 Attachment

10-00473, 1CMP-60-CR-15-1M Procedure Revision, Rev. 0 10-00731, Procedure 1

MSP-36.43-E Revision, Rev. 0 10-00687, Replacement of BV1 Vital Bus Alternate Source Regulating Transformers, Rev. 0 Modification Packages (* designates Modification and 10

CFR 50.59 screen sample) *ECP 06-0227, Add RWST Level Interlock to RS Pump Start, Rev. 1

ECP 07-0076, TDAFW Pump Steam Supply SOVs -Benchboard Control Switch Replacements, Rev. 1 *ECP 07-0145, Modification of U1 OS lines, Rev. 0 *ECP 08-0095, Unit 1 Emergency Diesel Generator K1 Relay Replacement, Rev. 1 *ECP 08-0134, Reclosing Reactor Trip Breaker Under Test, Rev. 0

ECP 08-0504, Unit 2 Joint Owners Group Motor Operated Valve Periodic Verification (JOG MOV PV) Program Implementation -2R14

ECP 09-03277, Recirculation Spray Heat Exchanger Inputs for MAAP Containment Analysis, Rev. 0 *ECP 10-0010, Unit 1 Containment Sump Discharge Flow Transmitter Replacement, Rev. 1 10080-N-584, Revision to Maximum Differential Pressure Across the Category I Motor Operated Valves In the Unit 2 Safety Injection (SI) System, Rev. 0 Calculations & Analysis 8700-DMC-1651, Containment Coatings Walkdown, Rev. 0 8700-DMC-1683, Required Height of OS Loop Seals, Rev. 0 10080-N-661, Torque Calculations for 2SIS-MOV867A, 2SIS-MOV867B, 2SIS-MOV867C, 2SIS-MOV867D, Rev. 12 10080-N-662, Torque Calculations for 2SIS-MOV869A and 2SIS-MOV869B, Rev. 8 10080-N-824, Recirculation Spray Heat Exchanger Inputs for MAAP Containment Analysis, Rev. 5 10080-US(B)-239, Assessment of Beaver Valley Unit 2 Containment Response for Design Basis Accidents for Containment Atmospheric Conversion Project, Rev. 5 12241-NP(B)-X70D, Pipe Stress Calculation -Reactor Coolant By-Pass -2 Inch Pipe Loop 21 -Reactor Containment Building, Rev. 3 12241-NP(B)-X70E, Pipe Stress Calculation -Reactor Coolant By-Pass -2 Inch Pipe Loop 22-Reactor Containment Building, Rev. 3 Corrective Action Reports 06-06553 07-28237

07-28510

08-38463

08-39206 (* denotes NRC identified during this inspection) 08-40329 10-78035 09-57719 10-78627* 10-71378 10-78665* 10-77661 10-78705* 10-77667 10-78734* Attachment

Drawings

8700-RE-0018F, Sht. 5, Wiring Diagram -120V AC Wiring Details, Rev. 8 8700-RE-0011 F, Sht. 2, Wiring Diagram -120 VAC DIST PNLS 7, 8, 9, 10, 11, 12, and 13, Rev. 36 8700-RE-21 BU, Elementary Diagram -Diesel Gen Engine Controls Sheet 2 of 4, Rev. 20 8700-RE-0021TZ, Elementary Diagram -Reactor Trip Switchgear 52/RTA, 52/RTB, 52/BYA, 52/BYB, Rev. 7 10080-E-6HT, Boron Injection Isolation Valves, Rev. 15 10080-E-11 LV, Train A Signal Isolators, Rev. 8 1 0080-RE-4AF, Sht. 1, Wiring Diagram Reactor Protection Rack Input Cabinet Train B, Rev. 12 10080-RE-9GK, Sht. 4, 480V MCC*2-E03 Aux Bldg, Rev. 25 1 0080-RM-0087 A, BV2 Flow Diagram Safety Injection Piping, Rev. 28 10080-RM-0087B, BV2 Flow Diagram Safety Injection Piping, Rev. 22 1 0080-RM-0406-001, Valve Oper. No Diagram Reactor Coolant System, Rev. 23 10080-RM-0413-002, Valve Diagram Quench Spray System, Rev. 19 12241-E-5DQ, Recirculation Spray Pump (2RSS*P21A), Rev. 13 Procedures 1

MSP-1.04-1, Reactor Protection System Train A Test, Rev. 47 1

MSP-9.05-1, Containment Sump Flow Measuring System Calibration, Rev. 9 1MSP-36A3-E, 1N 480 Volt Emergency Bus Degraded Voltage Relays 27-RN2100AB and 27-RN2100BC Test, Rev. 23 1

OM-9.1. E, Specific Instrumentation and Controls, Rev. 4 10M-9.3.C, Power Supply and Control Switch List, Rev. 6 1

OM-53A.1.E-0 (ISS1 C), Reactor Trip or Safety Injection, Rev. 11 10M-53A.1.ES-1.3 (ISS1C), Transfer to Cold Leg Recirculation, Rev. 7 1

OM-53E.1.SAG-3, Inject Into the RCS, Rev. 3 10M-36A.AH, Diesel Generator No.2 Start-Up and Shutdown, Rev. 11 1

OM-37.4.X, 480V Buses 1 G and 1 H Maintenance Support, Rev. 3 1

OM-37 A.Y, 480V Buses 1 E and 1 F Maintenance Support, Rev. 2 10M-38.5.B.2, Table 38-2 AC Distribution Panels Load List, Rev. 30 10M-45GA.AAA, Seismic Accelograph Operation, Rev. 5 1 OST -1.22, Manual ESF Actuation Circuitry Test, Rev. 5 10ST-13.7C, 1A Recirculation Spray Pump Auto Start Test, Rev. 6 10ST-36.2, Diesel Generator No.2 Monthly Test, Rev. 54 1

PMP-36EE-EG-1-2-1E, Emergency Diesel Generator Relay Cleaning and Inspection, Rev. 12 1/2-CMP-E-39-366, Station Battery Jumper Installation and Restoration, Rev. 7 2MSP-1.04-1, Reactor Protection System Train A Test, Rev. 44 2BVT 1.13.5, Recirculation Spray Pump Test, Rev. 15 2MSP-1.04-1, Reactor Protection System Train A Test, Rev. 44 2MSP-6.38-1, Reactor Coolant Temperature Loop 2RCS-T412 Delta T-Tavg Protection Channel I Calibration, Rev. 12 2MSP-13.05-1, 2QSS-L 104A, Refueling Water Storage Tank 2QSS-TK21 Level Loop Channell Calibration, Rev. 15 20M-6A.AH, Filling the RCS/RHS Piping with the Temporary Reactor Vessel Cover Installed, Rev. 0 Attachment

20M-13.4.AAB, Quench Spray System Trouble, Rev. 8 20M-24.4.ABM, AFW Pump SOV SS Closed, Rev. 0 20M-53A.1.E-0 (ISS1 C), Reactor Trip or Safety Injection, Rev. 8 20M-53E.1.SAG-6, Control Containment Conditions, Rev. 3 20M-53A.1.E-3 (ISS1 C), Steam Generator Tube Rupture, Rev. 15 20M-53A.1.ES-1.3 (ISS1 C), Transfer to Cold Leg Recirculation, Rev. 6 20ST -1.10, Cold Shutdown Valve Exercise Test, Rev. 33 20ST-10.3, Residual Heat Removal System Train A Valve Exercise, Rev. 24 20ST-13.3, Recirculation Spray Pump [2RSS*P21A] Dry Test, Rev. 12 20ST-13.3A, Recirculation Spray Pump [2RSS*P21A] Automatic Start Circuit Test, Rev. 3 20ST-24.4, Steam Driven Auxiliary Feed Pump [2FWE*P22] Quarterly Test, Rev. 65 20ST-47.3M, Containment Penetration and ASME Valve Test, Rev. 19

NOP-CC-2004, Design Interface Reviews and Evaluations, Rev. 7 Work Orders

200249586

200252480

200253464

200253465

200338908

200338912

200284820

200286520

200286513

200313752 Vendor Manuals 07.561-0023, Instructions and Parts Lists for Magnetic Flowmeter Instrumentation, Rev. F Miscellaneous 10BD-36A, Design Basis Document for Emergency Diesel Generators, Rev. 11 20BO-11, Design Basis Document for Safety Injection System, Rev. 12 1

SQS-36.2, Unit 1 Diesel Generator Operator Training Lesson Plan, Rev. 16 Licensed Operator Training Module 1

SQS-13.1, Containment Depressurization System, Rev. 1 Licensed Operator Training Module 2SQS-24.1, Feedwater Systems, Rev. 1 Preparation for

NRC 10CFR50.59/Modification Inspection Self-Assessment, dated 06/04/10 Technical Specifications, Beaver Valley Power Station Unit 1, Amendment 280 Technical Specifications, Beaver Valley Power Station Unit 2, Amendment 164 Attachment

ADAMS CFR CR DBA

DP DRS ECP EDG EPRI FENOC GL HX IP JOG LHSI MAAP MOV NEI NPSH OS RS RWST SGTR SI SOV TDAFW TS UFSAR A-5

LIST OF ACRONYMS

NRC "Document System Code of Federal Regulations Condition Report Design Basis Accident Differential Pressure Division of Reactor Safety Engineering Change Package Emergency Diesel Generator Electric Power Research Institute FirstEnergy Nuclear Operating Company Generic Letter Heat Exchanger Inspection Procedure Joint Owners Group Low Head Safety Injection Modular Accident Analysis Program Motor Operated Valve Nuclear Energy Institute Net Positive Suction Head Quench Spray Recirculation Spray Refueling Water Storage Tank Steam Generator Tube Rupture Safety Injection Solenoid Operated Valves Turbine Driven Auxiliary Feedwater Technical Specifications [[Licensing Basis Document" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Attachment]]