IR 05000277/2013010

IR 05000277-13-010 and 05000278-13-010; on 05/20/13 to 06/07/13; Peach Bottom Atomic Power Station (Pbaps), Units 2 and 3; Engineering Specialist Plant Modifications Inspection

ML13199A295
Person / Time
Site:	Peach Bottom
Issue date:	07/17/2013
From:	Paul Krohn Engineering Region 1 Branch 2
To:	Pacilio M Exelon Generation Co, Exelon Nuclear
References
IR-13-010
Download: ML13199A295 (20)

Text

UNITED STATES uly 17, 2013

SUBJECT:

PEACH BOTTOM ATOMIC POWER STATION - NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000277/2013010 AND 05000278/2013010

Dear Mr. Pacilio:

On June 7, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. The enclosed inspection report documents the inspection results, which were discussed on June 7, 2013, with Mr. Michael Massaro, Peach Bottom 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 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 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, 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,

/RA/

Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety

ML13199A295 Non-Sensitive Publicly Available SUNSI Review Sensitive Non-Publicly Available OFFICE RI/DRS RI/DRP RI/DRS NAME FArner MGray PKrohn DATE 07/08/13 7/12/13 7/17/13 Docket Nos. 50-277, 50-278 License Nos. DPR-44, DPR-56

Enclosure:

Inspection Report 05000277/2013010 and 05000278/2013010 w/Attachment: Supplemental Information

REGION I==

Docket Nos.: 50-277, 50-278 License Nos.: DPR-44, DPR-56 Report Nos.: 05000277/2013010 and 05000278/2013010 Licensee: Exelon Generation Company, LLC Facility: Peach Bottom Atomic Power Station, Units 2 and 3 Location: Delta, Pennsylvania Inspection Period: May 20 through June 7, 2013 Inspectors: F. Arner, Senior Reactor Inspector, Division of Reactor Safety (DRS),

Team Leader D. Orr, Senior Reactor Inspector, DRS J. Brand, Reactor Inspector, DRS Approved By: Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000277/2013010 and 05000278/2013010; 05/20/13 - 06/07/13; Peach Bottom Atomic

Power Station (PBAPS), Units 2 and 3; Engineering Specialist Plant Modifications Inspection.

This report covers a 2 week inspection of the evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three region based engineering inspectors. The NRCs 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.

ii

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications

(IP 71111.17)

.1 Evaluations of Changes, Tests, or Experiments (27 samples)

a. Inspection Scope

The team reviewed eight 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 Title 10 of the Code of Federal Regulations (10 CFR) Part 50.59 requirements. In addition, the team evaluated whether Exelon had been required to obtain U.S. Nuclear Regulatory Commission (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 (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 nineteen 10 CFR 50.59 screenings for which Exelon had concluded that a safety evaluation was not required to be performed. These reviews were performed to assess whether Exelon's threshold for performing safety evaluations was consistent with the requirements of 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 plant modifications inspection) not previously reviewed by NRC inspectors. The screenings and applicability determinations were selected based on the safety significance, risk significance, and complexity of the change to the facility.

In addition, the team compared Exelons 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 were identified.

.2.1 Repair of A Emergency Service Water Header Piping

a. Inspection Scope

The team reviewed an Engineering Change Request (ECR) 10-00293 that implemented a repair for a flaw in the 20-inch diameter A Emergency Service Water (ESW) header discharge piping in the Unit 2 high pressure service water (HPSW) room. The piping repair was required to address locations of low minimum pipe wall thickness identified during non-destructive testing of the HPSW piping. The repair consisted of a hot tap branch connection (metal plate with an isolation valve) welded over the flaw area.

Exelon evaluated the modification to ensure the design and licensing bases of the plant were not adversely affected by the engineering change.

The team reviewed the modification to verify that the design and licensing bases and performance capability of the HPSW system function had not been degraded. The team interviewed design engineers and reviewed post modification test results and associated maintenance work orders to confirm that the modification was appropriately implemented. The team also reviewed applicable corrective action issue reports (IRs)and performed a partial walkdown of the HPSW system to visually inspect the pipe repair. The 10 CFR 50.59 screening determination associated with the modification was also reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

.2.2 Emergency Diesel Generator Fuel Oil Day Tank Level Switch Settings

a. Inspection Scope

The team reviewed ECR 10-00337 that re-evaluated fuel oil day tank level switch setting values to determine the margin between the Emergency Diesel Generator (EDG) fuel oil day tank level switch setpoints and the TS requirements. Exelons evaluation was also performed in part to ensure proper setpoint values were implemented to minimize fuel oil transfer pump cycling. The evaluation was applicable for both Units 2 and 3. The ECR and associated calculation incorporated the use of ultra-low-sulfur-diesel (ULSD) fuel oil.

The team reviewed the modification and associated level instrumentation calculations and calibration procedures to confirm that the design and licensing bases and performance capability of the EDG fuel oil day tanks had not been degraded by the modification. The team interviewed Exelon design and system engineers, reviewed the modification package and reviewed vendor documents associated with the use of Ultra Low Sulfur Fuel (ULSF) oil to determine if the newly established level switch settings met the design and licensing bases requirements. The team reviewed Exelons initial evaluation and justification for the use of ULSD fuel oil performed in ECR PB 07-00073.

The team reviewed NRC Information Notice 2006-22, New Ultra Low Sulfur Diesel Fuel Oil could Adversely Impact Diesel Engine Performance, to evaluate whether Exelon had properly considered the impact of the ULSD fuel oil. The team reviewed post modification testing results and associated work orders for the modification to verify proper EDG operation. Additionally, the team walked down the EDGs and associated fuel oil day tanks to verify proper material condition. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

.2.3 Reactor Water Cleanup High Energy Line Break Analysis

a. Inspection Scope

The team reviewed ECR 12-00263 that performed a re-analysis of a High Energy Line Break (HELB) postulated in four areas of the reactor water cleanup (RWCU) system.

The analysis performed an evaluation for both the current licensed thermal power and the proposed extended power uprate (EPU) conditions. The four plant areas evaluated included:

Isolation Valve Compartment RWCU Pump Rooms RWCU Regenerative Heat Exchanger Room RWCU Non-Regenerative Heat Exchanger Room The team reviewed the modification to verify that the design and licensing bases and performance capability of the RWCU isolation system would not be impacted by the new HELB mass and energy release values for the postulated breaks. The team interviewed Exelon design and system engineers and reviewed the modification package to verify that the RWCU system isolation function would still meet the design and licensing bases requirements. The team reviewed documentation associated with applicable RWCU containment isolation motor-operated valves (MO-2/3-12-015 and MO-2/3-12-018)including design calculations and weak link analyses to verify the motor operators were qualified for the environmental conditions. Documents reviewed are listed in the

.

b. Findings

No findings were identified.

.2.4 Residual Heat Removal Injection Valve, MO-3-10-154A(B), Valve Control Logic

Modification

a. Inspection Scope

The team reviewed ECR 10-00363 that changed the closing logic for the Unit 3 residual heat removal (RHR) Loops A and B recirculation outer injection valves, MO-3-10-154A and MO-3-10-154B. These normally open motor-operated valves (MOVs) are located in the discharge lines from the A and B loop low pressure coolant injection (LPCI) pump to the reactor coolant system recirculation line and serve as the outboard isolation valves. The valves perform an active safety function in the open position and also perform an active safety function in the closed position during post-accident conditions to allow manual alignment for containment cooling. The modification replaced the 2-rotor switch with a 4-rotor switch and bypassed the torque switch stop signal until actuation of the closed limit switch. The modification was implemented to allow the use of the full capability of the valve motor to close the valve rather than be limited by the torque switch setting.

The team reviewed selected design inputs and attributes to ensure that they were consistent with the design and licensing bases. The team verified that the safety-related component qualification for the new switch was adequate. Additionally, the team evaluated the new design to ensure that it did not introduce any new failure modes for the valve which could impact the RHR system containment cooling design basis response. The team reviewed the post modification test plan and results to ensure the valve performance met the established acceptance criteria. Finally, the team interviewed the motor-operated valve engineer and design engineer to discuss the implementation of the modification. The 10 CFR 50.59 screening determination associated with the modification was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

.2.5 High Pressure Coolant Injection Pump Suction Valve, MO-2-23-058, Actuator

and Control Logic Modifications

a. Inspection Scope

The team reviewed ECRs 09-00175 and 09-00275 that revised the control logic for the normally closed Unit 2 MOV located in the supply line from the suppression pool to the high pressure coolant injection (HPCI) pump suction. The valve performs an active safety function in the closed position and is a containment isolation valve. The modifications were implemented to increase the motor operator overall gear ratio and improve the limit switch design to utilize rotors 3 and 4 in order to set up the MOV for limit switch control. These changes were installed to improve the motor capability through gearing, and allowed flexibility in valve setup such that the required thrust window is achievable without challenging the motor performance capability and structural margin of the MOV. The available operator torque for MO-2-23-058 was increased by changing its gear set and gear ratio.

The team assessed selected design inputs and attributes to ensure that they were consistent with the design and licensing bases. The team verified that the safety-related component qualification for the actuator modification was adequate. Additionally, the team evaluated the new design to ensure that it did not introduce any new failure modes for the valve which could impact the HPCI system design basis response. The team reviewed the post modification test plan and results to ensure valve performance met the established acceptance criteria for the new design. Finally, the team interviewed the MOV engineer and design engineer to discuss the implementation of the modification.

The 10 CFR 50.59 screening determination associated with the modification was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

.2.6 Design Power Feed to Station Blackout Control Power Transformer

a. Inspection Scope

The team reviewed ECR 10-00328 that modified the control power transformer (CPT) to the station blackout (SBO) alternate AC (AAC) power source. The previous CPT had a history of failures because its configuration was vulnerable to transient system conditions including ferroresonance. The replacement CPT was a grounded wye/grounded wye connection and the original CPT was a delta connection on the primary side, susceptible to ferroresonance. The modification evaluated, designed, and installed the new CPT which was powered from the 33kV SBO line.

The team reviewed the modification to verify that the design and licensing bases and performance capability of the SBO AAC source was not degraded by the modification.

The team assessed Exelons technical evaluations and design details, including installation specifications, and interviewed engineering personnel to determine whether the AAC would function in accordance with the modification's assumptions, and with design and licensing requirements. Drawings and procedures were reviewed to determine whether they were properly updated to reflect the post modification design and operation. The team also reviewed completed work orders to assess whether installation activities were performed as specified by the modification's design. The post modification results were reviewed to determine that the acceptance criteria had been met. In addition, the team walked down the AAC control power transformer and associated cable and hardware modifications to independently evaluate material conditions and configuration control with the approved design. Additionally, the 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the

.

b. Findings

No findings were identified.

.2.7 Item Equivalency Change for 4kV and 13kV Breaker Control Switches

a. Inspection Scope

The team reviewed ECR 11-00168 that evaluated new control switches as a replacement for existing control switches within safety-related 4kV breakers. The control switch replacement is also available for 13kV breakers but this voltage level has no safety-related application. Existing control switches were identified by the vendor to be under-rated for ampacity, were unserviceable, and had several industry related failures as documented in NRC Information Notice 97-08, Potential Failures of General Electric Magne-Blast Circuit Breaker Subcomponents, dated March 12, 1997. However, the team noted that Exelon had not experienced the failures described in the Information Notice. The control switch provides important control functions necessary for breaker operability. The vendor has recently provided a replacement control switch that was appropriately rated for the design current conditions.

The team reviewed the modification to verify that the design and licensing bases of the AC electrical system was not degraded by the breaker control switch replacements.

The team interviewed engineering staff and reviewed technical evaluations associated with the modification to determine whether the control switches would perform as required. The team reviewed electrical design evaluations to verify the electrical characteristics of the control switches were appropriately evaluated and justified. The team reviewed planned work orders and maintenance procedures to verify Exelon was timely in replacing the aging control switches with new switches as available, and that existing maintenance practices maintained the breaker operable with the old style control switch. The documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

.2.8 Offsite Power Transformer Load Tap Changer Contingency Voltage Drop Limits

a. Inspection Scope

The team reviewed ECR 11-00168 that revised abnormal operating procedures to ensure the operability of the offsite power sources during post trip conditions. The post trip contingency values were revised as a result of an operating experience review (OER) conducted by Exelon. The OER identified a weakness in electrical studies that did not consider the minimum voltage necessary to operate load tap changer (LTC)motors. Additionally, Exelon verified the availability of electric power to the LTC motors was not compromised by procedure steps that secured non-vital loads during grid emergency conditions.

The team reviewed the modification to verify that the design and licensing bases of the AC electrical system was not degraded by the operating procedure changes. The team interviewed engineering staff and reviewed technical evaluations associated with the abnormal operating procedure changes. The team verified that the procedure changes and design and license basis changes were accurately reflected in recent revisions.

Finally, the 10 CFR 50.59 applicability 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. Findings

No findings were identified.

.2.9 Residual Heat Removal Containment Spray Isolation Valve Motor Replacement

a. Inspection Scope

The team reviewed ECR 12-00415 that replaced an RHR containment spray isolation valve motor. The actuator motor was replaced as a corrective action after overheating symptoms were visually observed during an internal inspection of the magnesium rotor.

The internal inspection was a routine preventive maintenance activity for magnesium rotor valve actuator motors. The replacement motor had an aluminum rotor and additional physical and electrical differences that required evaluation and justification.

The motor control center bucket was also reworked with a replacement breaker, thermal overload heaters, and revised trip settings.

The team reviewed the modification to verify that the design and licensing bases and performance capability of RHR valve MO-2-10-031B was not degraded by the modification. The team interviewed engineering staff and reviewed technical evaluations associated with the modification to determine whether the motor operated valve would perform as required. The team reviewed electrical design evaluations to verify the electrical characteristics of the motor did not adversely impact the AC distribution system. The team reviewed work orders and the post modification testing to ensure MO-2-10-031B was properly returned to service and tested appropriately. Finally, the team reviewed design calculations, evaluations, and drawings to verify they were properly updated after the design modification. The documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

.2.10 Evaluation of Low Pressure Coolant Injection Capability during Residual Heat Removal

Suppression Pool Cooling Mode Alignment

a. Inspection Scope

The team reviewed ECR 08-00438 that revised the Units 2 and 3 Technical Specification Bases (TSB) to optimize when a subsystem of Low Pressure Coolant Injection (LPCI) is declared inoperable while a loop of the RHR system is aligned in the suppression pool cooling mode. The evaluation resulted in procedure changes which incorporated specific guidance on the effect on LPCI operability when suppression pool cooling valves are open during times when an EDG is out of service. This was performed because under certain conditions, a failure of an EDG to operate, while RHR is in the suppression pool cooling mode, could result in the LPCI mode of RHR not being capable of meeting its design bases requirements. The evaluation considered various configurations including the limiting loss-of-coolant/loss-of-offsite-power (LOCA/LOOP) licensing bases condition.

The team reviewed the ECR to verify that the design and licensing bases of the RHR system was not degraded by the procedure revisions. The team interviewed engineering staff and reviewed the associated technical evaluation for the procedure changes. The team verified that the procedure changes and design and licensing bases changes were accurately reflected in recent revisions. Finally, the 10 CFR 50.59 applicability determination associated with this change was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems (IP 71152)

a. Inspection Scope

The team reviewed a sample of corrective action documents 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 or completed corrective actions were appropriate. In addition, the team reviewed Condition Reports (CRs) written on issues identified during the inspection to verify adequate problem identification and incorporation of the issues into the corrective action system. The CRs reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA6 Meetings, including Exit

The team presented the preliminary inspection results to Mr. M. Massaro, Site Vice President, and other members of Exelons staff at a meeting on June 7, 2013. The team returned proprietary information reviewed during the inspection and verified that this report does not contain proprietary information.

ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

J. Chizever, Mechanical Design Supervisor

J. Coyle, Electrical Design Engineer

K. Cutler, Senior Electrical/I&C Design Engineer

W. Ford, System Engineer

K. Forney, Procurement Engineer

J. Futcher, Electrical Design Engineer

M. Hoffman, Electrical Design Engineer

J. Laverde, Mechanical Design Engineer

T. Moore, Site Engineering Director

NRC Personnel

S. Hansell, Senior Resident Inspector

ITEMS OPENED, CLOSED AND DISCUSSED

None.