IR 05000277/2011002

IR 05000277-11-002 & 05000278-11-002, on 01-01-11 - 03-31-11, Peach Bottom Atomic Power Station, Integrated Inspection Report

ML111260700
Person / Time
Site:	Peach Bottom
Issue date:	05/06/2011
From:	Paul Krohn Reactor Projects Region 1 Branch 4
To:	Pacilio M Exelon Nuclear, Exelon Generation Co
Krohn P
References
IR-11-002
Download: ML111260700 (46)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD KlNG OF PRUSSlA. PA 19406-1415 May 6, 2017 Mr. Michael Senior Vice President, Exelon Generation Company, LLC President and Chief Nuclear Officer, Exelon Nuclear 4300 Winfield Road Warrenville. lL 60555 SUBJECT: PEACH BorroM AToMlc PowER srATtoN - NRc TNTEGRATED I NS PECTI O N RE PORT 0500027 7 t20 1 1 002 AN D 0500 027 8t20 1 1 002

Dear Mr. Pacilio:

On March 31, 2011, the U. S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. The enclosed integrated inspection report documents the inspection results, which were discussed on April 29,2011, with Mr. Thomas Dougherty 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.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, one finding of significance was identified. This finding was determined to involve a violation of NRC requirements. Additionally, two licensee-identified violations which were determined to be of very low safety significance are listed in this report.

However, because of the very low safety significance and because the findings have been entered into your correction action program (CAP), the NRC is treating the findings as a non-cited violations (NCVs), consistent with Section 2.3.2 of the NRC's Enforcement Policy.

lf you contest any of the NCVs in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the RegionalAdministrator, Region l; the Director, Office of Enforcement, U. S. NRC, Washington, DC 20555-0001; and the NRC Senior Resident Inspector at the PBAPS. lf you disagree with the cross-cutting aspect to the finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I and the NRC Senior Resident lnspector at PBAPS. The information you provide will be considered in accordance with lnspection Manual Chapter (lMC) 0305. ln accordance with Title 10 of the Code of Federal Regulations (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 Website at http:l/www.nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).

Uaru.

Sincerely, Paul G. Krohn, Chief Projects Branch 4 Division of Reactor Projects Docket Nos.: 50-277, 50-278 License Nos.: DPR-44, DPR-56

Enclosure:

lnspection Report 0500027712011002and 0500027812011002 w/Attachment: Supplemental Information

REGION I Docket Nos.: 50-277,50-278 License Nos.: DPR-44, DPR-56 Report No.: 05000277 l 201 1 002 a nd 05000 27 81201 1 002 Licensee: Exelon Generation Company, LLC Facility: Peach Bottom Atomic Power Station, Units 2 and 3 Location: Delta, Pennsylvania Dates: January 1,2011 through March 31,2011 Inspectors: F. Bower, Senior Resident Inspector A. Ziedonis, Resident lnspector C. Crisden, Emergency Preparedness Specialist J. D'Antonio, Senior Operations Engineer T. Fish, Senior Operations Engineer S. Hammann, Senior Health Physicist J. Schoppy, Senior Reactor Inspector Approved by: Paul G. Krohn, Chief Reactor Projects Branch 4 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

lR 0500027712011002, 0500027812011Q02; 0110112011 - 031311201 1; Peach Bottom Atomic

Power Station (PBAPS), Units 2 and 3; Other Activities.

The report covered a three-month period of inspection by resident inspectors and announced inspections by a regional emergency preparedness specialist, two senior operations engineers, a senior health physicist, and a senior reactor inspector. One self-revealing finding was identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using IMC 0609, "significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.

Cross-cutting aspects associated with findings are determined using IMC 0310, "Components Within The Cross-Cutting Areas," dated February 2010. 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.

Cornerstones: Barrier Integrity

.

Green.

A Green self-revealing NCV of Technical Specification (TS) 5.4.1 "Procedures" was identified, because PBAPS's procedures for refueling equipment operation and core alterations were inadequate to prevent a fuel bundle from contacting a core spray inspection (CSl) submarine device while the fuel bundle was being transported from the core to the spent fuel pool (SPF). In particular, system operating (SO) procedure 18.1.A-2, "Operation of Refueling Platform," and fuel handling (FH) procedure 6C, "Core Component - Core Transfers," did not provide sufficient procedure steps, precautions, or human performance tools to prevent contact while the refueling platform was operated in the automatic mode and when core components were in close proximity to obstructions and interferences.

The inspectors determined that the finding was more than minor because the finding was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone's objective to provide reasonable assurance that physical design barriers (i.e., fuel cladding) protect the public from radionuclide releases caused by accidents or events. Although no fuel damage occurred during this event, the inadequate procedure resulted in a FH event that could have impacted the cladding and affected the cornerstone's objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. IMC 0609, "SDP," Attachment 0609.04, "Phase 1-lnitial Screening and Characterization of Findings," was used to evaluate the significance of the finding.

Attachment 0609.04, Table 4a, was used to evaluate the impact of the finding on fuel clad integrity. Appendix G was considered for the evaluation, but was not used because it does not directly address fuel clad integrity. Based on the results of fuel sipping done in February 2011, PBAPS concluded that there was no damage to the clad integrity of the impacted fuel bundle that was permanently discharged to the SFP. Since the finding did not affect SFP cooling or inventory and since there was no damage to fuel clad integrity from the impact with the CSI submarine, the finding was determined to be of very low safety significance (Green).

The finding has a cross-cutting aspect in Human Error Prevention Techniques in the Work Practices component of the Human Performance area. Specifically, PBAPS FH procedures did not require human error prevention techniques that were commensurate with the risk of moving fuel in close proximity to obstructions and interferences. (Section 4oA5.1)tH.a(a)l Other Findinqs Two violations of very low safety significance, which were identified by the licensee, have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's CAP. These violations and the licensee's corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summarv of Plant Status Unit 2 began the inspection period at 100 percent rated thermal power (RTP) where it generally remained until the end of the inspection period, except for brief periods to support planned testing, maintenance, and rod pattern adjustments.

Unit 3 began the inspection period at 100 percent RTP where it generally remained until the end of the inspection period, except for brief periods to support planned testing, maintenance, and rod pattern adjustments.

REACTORSAFETY Cornerstones: Initiating Events, Mitigating Systems, and Barrier lntegrity

1R04 Equipment Aliqnment (71111.04 - 5 Samples)

.1 Partial Walkdown (71111.04Q - 4 Samples)

a. Inspection Scope

The inspectors performed a partial walkdown of four systems to verify the operability of redundant or diverse trains and components when safety-related equipment was inoperable. The inspectors performed walkdowns to identify any discrepancies that could impact the function of the system and potentially increase risk. The inspectors reviewed selected applicable operations procedures, walked down system components, and verified that selected breakers, valves, and support equipment were in the correct position to support system operation. Documents reviewed during the inspection are listed in the Attachment. The four systems reviewed were:

o Unit 3 high pressure coolant injection (HPCI) during reactor core isolation cooling (RCIC) planned maintenance; o Unit 3 'A'train of residual heat removal (RHR) during 'B'train planned maintenance; o Unit 2 'A' train of core spray (CS) during 'B' train planned maintenance; and

.

Unit 3 'B'train of RHR during 'A'train planned maintenance.

b.

Findinqs No findings were identified.

.2 Complete Walkdown (71111.04S - 1 Sample)

a.

lnspection Scope The inspectors performed a complete walkdown of the accessible portions of the Unit 2

'B' train of the CS system to verify adequate alignment of the system to successfully perform its safety function, to satisfy TS 3.5.1 operability, and to assess general material condition of the system in the plant. lnspector walkdowns were performed in all accessible portions of the plant during full power operation, including the main control room panels. The inspectors reviewed system drawings and operating procedures to verify that the system alignment was properly translated into procedures and drawings.

The inspectors discussed system operation with the plant operators, and reviewed issue reports to verify that Unit 2 CS system 'B'train issues were properly being identified, evaluated, and corrected. Documents reviewed are listed in the Attachment.

b. Findinqs No findings were identified.

1R05 Fire Protection (71111.05Q - 6 Samples)

.1 Fire Protection - Tours (5 Samples)

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of firefighting equipment. The inspectors reviewed areas to assess whether PBAPS had implemented the Peach Bottom Fire Protection Plan (FPP) and adequately: controlled combustibles and ignition sources within the plant; maintained fire detection and suppression capability; and maintained the material condition of passive fire protection features. For the areas inspected, the inspectors also verified that PBAPS had followed the Technical Requirements Manual (TRM) and the FPP when compensatory measures were implemented for out-of-service (OOS),degraded or inoperable fire protection equipment, systems, or features. The inspectors verified: that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient combustible materials were managed in accordance with plant procedures; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition.

Documents reviewed during the inspection are listed in the Attachment. The inspectors toured the following areas:

.

Unit 3 'A' and 'C' RHR rooms - elevations 1 16' and 91'-6" (Fire Zones 1 1 and 12A);

.

Unit 2'A' and 'C' CSI and pump rooms - elevations 116'and 91'-6" (Fire Zones 5A, 58 and 5F);o Unit 3 'B' and 'D' RHR rooms - elevations 1 16' and 91 '-6" (Fire Zones 9 and 10);

.

Hydrogen cylinder storage (Fire Zone 150); and o Unit 2 radwaste building, reactor building closed-cooling water (RBCCW) room -

elevation 116'-0'.

.2 Fire Drill (1 Sample)

a. Inspection Scope

On March 16, the inspectors observed the performance of a fire drill scenario in the Unit 3 turbine building, 166' elevation, reactor feed pump / chiller area (Fire Zone 78L). The inspectors observed the drill to determine the readiness of the plant fire brigade to respond and combat fires. The inspectors focused the inspection of the fire brigade response, donning of the protective gear, fire brigade leader command and control, radio communication between the fire brigade leader and main control room, execution of the "two-in, two-out" approach, conformance with the fire drill scenario, execution of the drill objectives, and returning of firefighting equipment to a state of readiness.

The inspectors observed the post-drill critique to determine whether weaknesses and/or failures were appropriately identified, thoroughly and openly discussed in a self-critical manner, and that appropriate training and learning opportunities were identified and discussed. The inspectors also verified that issues discussed at the post-drill critique were appropriately documented to develop corrective actions for future training.

The inspectors verified that RT-F-101-922-2, "Fire Drill," was completed to record the fire drill scenario that was used, measure performance of the drill objectives, and capture the critique results. Documents reviewed are listed in the Attachment.

b. Findinqs No findings were identified.

1R06 lnternal Flood Protection (71111.06 - 1 Sample)

Underqround Cables (1 Sample - Underground Cables)

a. Inspection Scope

The Exelon Nuclear Cable Condition Monitoring Program is controlled under procedure ER-AA-3003, "Cable Condition Monitoring Program." The annual preventive maintenance inspection of all manholes containing safety-related and Maintenance Rule (MR) scoped cables was begun during this inspection period (work order (WO)

R1174133). From this inspection population, the inspectors selected three manholes (MH-40, 164, and 168) containing underground safety-related cables and one manhole (MH-004) containing underground cables within the scope of the MR as an internal flood protection measures sample for review. The inspectors directly observed the interior of the subject manholes and the associated cabling after the covers had been removed to determine whether cables in each of the four manholes inspected were submerged. The inspectors reviewed the work instructions to verify that PBAPS's inspections verify through direct observation: whether the cables in manholes are submerged in water; that the cables and/or splices and their supports are not damaged or degraded; and that the manhole drainage system, if installed, is functioning properly. The inspectors verified that issue reports (lRs) were being initiated for identified discrepancies and were entered into the CAP. Documents reviewed during the inspection are listed in the Attachment.

b. Findinss No new findings were identified since NCVs 05000277 and 27812009005-01 were previously issued.

1R07 Heat Sink Performance (71111.07 - 1 Sample)

a. Inspection Scope

The inspectors selected for review the thermal performance testing of the Unit 2 'A' RHR room cooler as one annual sample. This safety-related water (emergency service water (ESW)) to air heat exchanger (HX) is needed to provide the ventilation flow and cooling ihat assures operability of ine pump room's engineered safeguards equipment and the associated auxiliary equipment. To verify the readiness and availability of the fan cooling unit, the inspectors reviewed the data collected by RT-l-033-631-2, "RHR Room Cooler ESW Heat Transfer Test," for any obvious problems or errors. The inspectors independently verified that the test data was correctly transferred into the HX performance computer model and verified that the test acceptance criteria were met.

The inspectors also verified that the acceptance criteria were supported by the design basis calculation, PM-0958, "RHRyCS Pump Room Temperatures (post-loss-of-coolant accident) for 95' F River Temperature," Revision 2. Additionally, the inspectors reviewed selected portions of the "Balance HX Utility Theoretical and Verification Manual," that supports the HX performance computer model used by PBAPS.

Documents reviewed during the inspection are listed in the Attachment.

b.

Findinqs No findings were identified.

1R1 1 Licensed Operator Requalification Proqram (71111.11 - 3 Samples)

.1 Resident Inspector Quarterlv Review (71111.11Q - 1 Sample)

a. Inspection Scope

On January 31,2A11, the inspectors observed a simulator-based licensed operator evaluation, during requalification training, to assess licensed operator performance and the evaluator's post-scenario critique. The inspectors evaluated crew performance in the areas of:

.

Clarity and formality of communications;

.

Ability to take timely actions; o Prioritization, interpretation, and verification of alarms; o Procedure usage;

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Timely control board manipulations with a focus on high-risk operator actions; o Shift supervisor command and control, including identification and implementation of TSs, event classification and emergency response actions; and

.

Group dynamics involved in crew performance.

The inspectors verified that any crew performance issues and weaknesses were discussed in the post-scenario critique. The inspectors also verified simulator physical fidelity, to ensure that the simulator arrangement closely paralleled the main control room. Documents reviewed during the inspection are listed in the Attachment. These activities constituted one quarterly licensed operator requalification training program inspection sample.

b.

Findinqs No findings were identified.

I

.2 Biennial Limited Senior Reactor Operator (LSRO) Requalification Proqram (71111.118 -

1 Sample)a. Inspection Scooe On March 7, 2011, one NRC region-based inspector conducted an in-office review of the results of licensee-administered comprehensive written exams for the LSRO Requalification Program for 2010. The inspection assessed whether pass rates were consistent with the guidance of NRC Manual Chapter 0609, Appendix l, "Operator Requalification Human Performance SDP." The inspector verified that:

.

lndividual pass rates on the written exam were greater than or equal to 80 percent.

(Pass rate was 80 percent); and r Two individuals who failed the original written exam passed their remediation exam.

b. Findinqs No findings were identified.

.3 Biennial Licensed Operator Requalification Prooram (71111.118 - 1 Sample)

a. Inspection Scope

The following inspection activities were performed using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 9, Supplement 1, lnspection Procedure (lP) Attachment 71 111.11, "Licensed Operator Requalification Program," Appendix A, "Checklist for Evaluating Facility Testing Material," and Appendix B, "Suggested Interview Topics."

A review was conducted of recent operating history documentation found in inspection reports, licensee event reports (LERs), and the licensee's CAP. The inspectors also reviewed specific events from the licensee's CAP which indicated possible training deficiencies, to verify that they had been appropriately addressed. The senior resident inspector was also consulted for insights regarding licensed operators' performance.

These reviews did not detect any operational events that were indicative of possible training deficiencies.

The operating tests for the week of March 7, 2011, were reviewed for quality and performance.

On March 18, 2011, the results of the annual operating tests for year 2011 were reviewed to determine if pass fail rates were consistent with the guidance of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 9, Supplement 1, and NRC Manual Chapter 0609, Appendix l, "Operator Requalification Human Performance SDP." The review verified the following:

o Individual pass rates on the dynamic simulator test were greater than 80 percent.

(Pass rate was 98.4 percent);

.

Crew pass rates were greater than 80 percent. (Pass rate was 100 percent);

.

Individual pass rates on the job performance measures (JPM) of the operating examination were greater than 80 percent. (Pass rate was 100 percent); and o More than 75 percent of the individuals passed all portions of the examination.

(98.4 percent of the individuals passed all portions of the examination).

Observations were made of the dynamic simulator examinations and JPMs administered during the week of March 7,2011. These observations included facility evaluations of crew and individual performance during the dynamic simulator examinations and individual performance of seven JPMs.

The remediation plans for two written failures (the comprehensive written was administered in February and March of 2010) were reviewed to assess the etfectiveness of the remedial training.

Two license reactivation records were reviewed to ensure that 10 CFR 55.53 license conditions and applicable program requirements were met.

Operators, instructors, and training/operation's management were interviewed for feedback on their training program and the quality of training received.

Simutator performance and fidelity were reviewed for conformance to the reference plant control room.

A sample of records for requalification training attendance, program feedback, reporting, and medical examinations were reviewed for compliance with license conditions, including NRC regulations.

b. Findinqs No findings were identified.

1R12 Maintenance Effectiveness (71111J2Q - 2 Samples)

a. Inspection Scope

The inspectors evaluated PBAPS's work practices and follow-up corrective actions for safety-related structures, systems, and components (SSC's) and identified issues to assess the effectiveness of PBAPS's maintenance activities. The inspectors reviewed the performance history of SSCs and assessed PBAPS's extent-of-condition (EOC)determinations for those issues with potential common cause or generic implications to evaluate the adequacy of the PBAPS's corrective actions. The inspectors assessed PBAPS's problem identification and resolution (Pl&R) actions for these issues to evaluate whether PBAPS had appropriately monitored, evaluated, and dispositioned the issues in accordance with Exelon procedures, including ER-AA-3l0, "lmplementation of the MR," and the requirements of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance." ln addition, the inspectors reviewed selected SSC classifications, performance criteria and goals, and PBAPS's corrective actions that were taken or planned, to evaluate whether the actions were reasonable and appropriate.

Documents reviewed during the inspection are listed in the Attachment. The inspectors performed the following two samples:

.

Unit 2 - Primary Containment lsolation Valves MR (a)(1) Determination (lR 1165384); and

.

Manhole Water Intrusion System Deficiencies (lR 1 179383).

b.

Findinqs No findings were identified.

1R13 Maintenance Risk Assessments and Emerqent Work Control (71111.13 - 7 Samples)

a. Inspection Scope

The inspectors evaluated PBAPS's implementation of the Maintenance Risk Program with respect to the effectiveness of risk assessments performed for maintenance activities that were conducted on SSC's. The inspectors also verified that PBAPS managed the risk in accordance with 10 CFR Part 50.65(a)( ) and procedure WC-AA-101, "On-line Work Control Process." The inspectors evaluated whether PBAPS had taken the necessary steps to plan and control emergent work activities and to manage overall plant risk. The inspectors selectively reviewed PBAPS's use of the online risk monitoring software and daily work schedules. Documents reviewed during the inspection are listed in the Attachment. The activities selected were based on plant maintenance schedules and systems that contributed to risk. The inspectors completed seven evaluations of maintenance activities on the following:

.

Unit 3 essential bus E-13 overcurrent relay replacement (WO 1151273);

.

Unit 3 reactor protection system (RPS) motor-generator (MG) set voltage adjustments (troubleshooting, rework and test (TRT) 11-02);r Unit 2 HPCI maintenance (Clearan ce 1QQ02255) concurrent with breaker E-342 maintenance (Clearance 1 0002265);r Unit 3 RCIC unavailability and yellow on-line risk condition due to planned maintenance (Clearance 1 1000074);

.

Unit 3 'B'train of RHR unavailability and yellow on-line risk condition due to planned maintenance (Clearance 1 0002305);o Risk management actions associated with Unit 2'B' CS planned maintenance (Clearance 10001 1446); and r Unit 2 post-accident monitoring (PAM) power supply (lRs 1 170369 and 1 170050).

b.

Findinqs No findings were identified.

1R15 Operabilitv Evaluations (71111.15 - 7 Samples)

a. Inspection Scope

The inspectors reviewed seven issues to assess the technical adequacy of the operability determinations, the use and control of compensatory measures, and compliance with the licensing and design bases. Associated adverse condition monitoring plans (ACMPs), engineering technicalevaluations, and operational and technical decision making (OTDM) documents were also reviewed. The inspectors verified these processes were performed in accordance with the applicable administrative procedures and were consistent with NRC guidance. Specifically, the i nspectors referenced proced u re OP-AA- 1 08- 1 1 5, "Operability Determ inations, " and NRC IMC Part 9900, "Operability Determinations & Functionality Assessments for Resolutions of Degraded or Nonconforming Conditions Adverse to Quality or Safety."

The inspectors also used TSs, TRM, Updated Final Safety Analysis Report (UFSAR),and associated design basis documents (DBDs) as references during these reviews.

Documents reviewed during the inspection are listed in the Attachment. The following degraded equipment issues were reviewed:

r Switch # 3G3 increase heating on 'C' phase ball side finger (lR 1 164998);

.

OTDM for independent spent fuel storage installation (lSFSl) Cask #50 seal weld repair plan (lR 1129931);

.

3 'A' RHR HX - RHR (shell) to high pressure service water (HPSW) (tube)leak (lR 694879);

.

E-4 emergency diesel generator (EDG) governor actuator speed knob not set to value required by surveillance test (ST) (lR 1174526);

.

OD 11-02 for MO-26A leakage (lR 1178455);

.

OD 11-01 for General Electric safety communication 11-01 impact on Unit 2 marathon control blades (lR 1177911); and o Part 21 for Fairbanks Morse engine bearings (lR 1 172280).

b. Findinqs No findings were identified.

1 R18 Plant Modifications (71111.18 - 2 Samples)

.1 Permanent Modifications (1 Sample)

a. lnspection Scope The inspectors reviewed one permanent modification to verify that modification implementation did not place the plant in an unsafe condition. The review was also conducted to verify that the design bases, licensing bases, and performance capability of risk significant SSCs had not been degraded as a result of these modifications. The inspectors verified the modified equipment alignment through control room instrumentation observations; UFSAR, drawings, procedures, and WO reviews; staff interviews; and plant walkdowns of accessible equipment. Documents reviewed are listed in the Attachment. The following engineering change request (ECR) for a permanent modification was reviewed:

.

ECR 11-00062, ECR Required to Revise Calculation 49-481F (Standby Liquid Control Tanks and Pump Suctions Seismic Analysis).

b. Findinqs No findings were identified.

.2 Temporarv Modifications (1 Sample)

a. Inspection Scope

The inspectors reviewed the following temporary modification to ensure that it did not adversely affect the availability, reliability, or functional capability of any risk significant SSCs, and to verify that modification implei"nentation did not place the plant in an unsafe condition. The inspectors reviewed the applicable ECR, supporting documentation, and discussed the modification with engineering and maintenance, and operations personnel. The control of the modifications was compared to the regulatory requirements, regulatory guidance documents for on-line leak repairs, industry standards, and PBAPS procedural requirements. The inspectors also verified that the leak repair plan was consistent with the modification documentation, and that the drawings and the post-installation testing was adequate. Documents reviewed during the inspection are listed in the Attachment.

ECR 10-00405. Pressure Seal Steam Leak.

b.

Findinqs No findings were identified.

1R19 Post-Maintenance Testino (71111.19 - 6 Samples)

a.

lnspection Scope The inspectors reviewed completed test records or observed selected post-maintenance testing (PMT) activities. The inspectors verified whether the tests were performed in accordance with the approved procedures or instructions and assessed the adequacy of the test methodology based on the scope of maintenance work performed. ln addition, the inspectors assessed the test acceptance criteria to evaluate whether the test demonstrated that components satisfied the applicable design and licensing bases and the TS requirements. The inspectors reviewed the recorded test data to verify that the acceptance criteria were satisfied. Documents reviewed during the inspection are listed in the Attachment. The inspectors reviewed six PMTs performed in conjunction with the following maintenance activities:

o Unit 3 'C' HPSW pump, valve and flow and functional in-service test (lST) following pump replacement and Pl-3-32-3381A (C0235988 and M1789109);

.

Standby gas treatment system (SBGTS) filter train 'B'testing following planned maintenance WO R1 1 53751 );

.

Unit 2'B'RBCCW HX maintenance (WO R0930987);o Control rod (CR) stroke speed timing and CR scram timing on Unit 3 CR 46-39 after corrective maintenance to rod drift (WO C0236428-05);

.

E-3 EDG standby lube oil pump motor starter replacement (WO M1792292); and o Unit 2 RCIC pump, valve, flow and unit cooler functional and IST following testing and maintenance onMO-2-12-018 (WO R0931306).

b. Findinss No findings were identified.

1R22 Surveillance Testinq (71111.22 - 8 Samples)

a. Inspection Scope

(6 Routine Surveillances; 1 RCS Leak Detection; and 1 IST Sample)

The inspectors reviewed or observed selected portions of the following STs, and compared test data with established acceptance criteria to verify the systems demonstrated the capability of performing the intended safety functions. The inspectors also verified that the systems and components maintained operational readiness, met applicable TS requirements, and were capable of performing design basis functions.

Documents reviewed during the inspection are listed in the Attachment. The eight STs reviewed or observed included:

.

5T-0-020-560-213, Units 2 & 3 - Reactor Coolant Leakage Test (RCL) [1 RCS Leakage Samplel; ST-O-010-301-3, 'A' RHR Pump, Valve, Flow, and Unit Cooler Functional and lST, performed 01111111 [1 IST Sample];

ST-O-014-301-3, CS LOOP'A'Pump, Valve, Flow, and Cooler Functional and lST, performed 01119111;

.

M-01 8-107 , Control of Fuel Bundle Vacuum Sipping, performed 01110111 - 02104111;

.

ST-M-09A-601-2, SBGTS Filter Train '8,' performed 02119111;

.

Sl2M-60F-RT1 1-A2M2, Response Time Test of Main Steam Line High Radiation Scram Channels, performed Q3104111;

.

ST-O-052-703-2, E-3 Diesel Generator 24Hour Endurance Test, performed 03/10 -

===1112011; and

.

ST-O-023-350-2, HPCI Valve Alignment and Filled and Vented Verification, performed 0311512011, 0311712011, and 03/1 812011.

b. Findinqs No findings were identified.

Emergency Preparedness (EP)1EP2 Alert and Notification Svstem (ANS) Evaluation (71114.02 - 1 Sample)

Inspection Scope A review of the Peach Bottom ANS was conducted to assess current maintenance and testing practices. The inspectors reviewed ANS maintenance and testing procedures, maintenance and test records, and the updated Peach Bottom ANS design report to ensure Exelon's compliance with design report commitments for system maintenance and testing. A sample of condition reports (CRs) pertaining to the ANS was reviewed for causes, trends, and corrective actions. During the inspection, the inspectors interviewed the ANS System Manager to discuss system performance and upgrades. The inspection was conducted in accordance with NRC lP 71114, Attachment 2. Planning Standard, 10 CFR 50.47(bX5) and the related requirements of 10 CFR 50, Appendix E, were used as reference criteria.

Findinqs No findings were identified.

1EP3 Emerqencv Response Orqanization (ERO) Staffinq and Auqmentation Svstem (71114.03 - 1 Sample)

a. Inspection Scope

The inspectors conducted a review of Peach Bottom's ERO augmentation staffing requirements and the process for notifying and augmenting the ERO. This was performed to ensure the readiness of key licensee staff to respond to an emergency event and to ensure Exelon's ability to activate their emergency facilities in a timely manner. The inspectors reviewed the Peach Bottom Emergency Plan, Peach Bottom ERO duty roster, station augmentation reports, and CRs related to the ERO staffing augmentation system. The inspectors also reviewed a sampling of ERO responders training records to ensure training and qualifications were upto-date. During emergency events and exercises, the Emergency Offsite Facility is staffed by Exelon Mid-Attantic corporate staff. A review of the corporate ERO duty roster, augmentation results, and training records was also conducted. The inspection was conducted in accordance with NRC lP 71114, Attachment 3. Planning Standard, 10 CFR 50.47(b)(2) and related requirements of 10 CFR 50, Appendix E, were used as reference criteria.

b. Findinqs No findings were identified.

1EP4 Emerqencv Action Level (EAL) and Emerqencv Plan Chanqes (71114.04 - 1 Sample)

a. Inspection Scope

Since the last NRC inspection of this program area, Exelon implemented various changes to their standard Emergency Plan, the Peach Bottom Emergency Plan Annex, and implementing procedures. Exelon had determined that, in accordance with 10 CFR 50.54(q), any change made to the Plan, and its lower-tier implementing procedures, had not resulted in any decrease in effectiveness of the Plan, and that the revised Plan continued to meet the standards in 50.47(b) and the requirements of 10 CFR 50, Appendix E. The inspectors reviewed all EAL changes, including changes to the security EALs as endorsed by Nuclear Energy Institute (NEl) 99-02, Revision 5. A sample of emergency plan changes, including the changes to lower-tier emergency plan implementing procedures, were evaluated for any potential decreases in effectiveness of the Standard Emergency Plan and the Peach Bottom Emergency Plan Annex.

However, this review by the inspectors was not documented in an NRC Safety Evaluation Report (SER) and does not constitute formal NRC approval of the changes.

Therefore, these changes remain subject to future NRC inspection in their entirety. The inspection was conducted in accordance with NRC lP 71114, Attachment 4. The requirements in 10 CFR 50.5a(q) were used as reference criteria.

b. Findinqs No findings were identified.

1EP5 Correction of EP Weaknesses (71114.05 - 1 Sample)a. Inspection Scooe The inspectors reviewed a sampling of self-assessment procedures and reports to assess Exelon's ability to evaluate their Peach Bottom EP Performance and program.

The inspectors also reviewed drill reports, a 10 CFR 50.54(t) audit, and an EP performance report. A sampling of CRs initiated by Exelon at Peach Bottom from drills and audits between June 2010 and February 2011were reviewed. The inspectors also conducted a similar review for the Mid-Atlantic EP corporate functions. This inspection was conducted in accordance with NRC lP 71114, Attachment 5, Planning Standard, 10 CFR 50.47(bX14) and the related requirements of 10 CFR 50, Appendix E, were used as reference criteria.

b. Findinqs No findings were identified.

lEPO Drill Evaluation (71114.06 - 1 Sample)

.1 Simulator Traininq Observation (1 Simulator Training Sample)

a. Inspection Scope

During swing shift hours on January 31, 2011, the inspectors observed the classification and notification aspects of a licensed operator requalification training examination scenario in the PBAPS simulator. The scenario was conducted, in part, to provide drill and exercise performance (DEP) opportunities for the DEP performance indicator (Pl).

The inspectors reviewed the conduct of the simulator exercise to identify any weaknesses and deficiencies in classification and notification activities. The inspectors observed the evaluation, classification, and notification of the simulated events to ensure they were accurate, timely, and were done in accordance with EP-AA-1007 , "Exelon Nuclear Radiological Emergency Plan Annex for PBAPS." The inspectors verified that the drill evaluators correctly counted the drill's contribution in the calculation of the DEP Pl. The inspectors verified that training evaluators captured the results for the DEP Pl.

The inspectors also verified that any weaknesses or deficiencies were captured and discussed during the critique of the training exercise, in order to properly identify and correct any weaknesses. Documents reviewed during the inspection are listed in the

. The following simulated event was classified during this training exercise:

.

MS3 - Site Area Emergency, Failure of RPS Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a RPS Has Been Exceeded and Manual Scram Was NOT Successful.

b. Findinos No findings were identified.

4OA1 Performance Indicator Verification (71151- 9 Samples)

.1 Mitiqatinq Svstems Pls (71151- 6 Samples)

a.

lnspection Scope The inspectors reviewed a selected sample of the PBAPS's information submitted for the six Mitigating Systems Pls listed below to assess the accuracy and completeness of the data reported to the NRC for these Pls. The Pl definitions and the guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Revision 6, and Exelon procedure LS-AA-2200, "Mitigating System Performance Index Data Acquisition and Reporting," Revision 3, were used to verify that procedure and reporting requirements were met. The inspectors reviewed raw Pl data collected from October 2009 through September 2010 and compared graphical representations from the applicable Pl reports to the raw data to verify the data was included in the report. The inspectors also examined a selected sample of operations logs, LERs, CAP records, equipment clearances, and MR data to verify the Pl data was appropriately captured for inclusion into the Pl report and that the individual Pls were correctly calculated. Documents reviewed during the inspection are listed in the Attachment.

Units 2 and 3

.

Unplanned SCRAMS per 7,000 Critical Hours;

.

Unplanned SCRAMS with Complications; and

.

Unplanned Power Changes per 7,000 Critical Hours.

b. Findinos No findings were identified.

.2 Review of Peach Bottom's EP Pls (71151- 3 Samples)

a. lnspection Scope The inspectors reviewed data for Peach Bottom's EP Pls, which are:

(1) Drill and Exercise Performance

(2) ERO Drill Participation; and

(3) ANS Reliability. The inspectors reviewed the Pl data and its supporting documentation from the second quarter of 2010 through the fourth quarter of 2010 to verify the accuracy of the reported data. The review of these Pls was conducted in accordance with NRC lP 71151, using the acceptance criteria documented in NEI 99-02, "Regulatory Assessment Performance lndicator Guidelines," Revision 6.

b. Findinqs No findings were identified.

4c42 ldentification and Resolution of Problems (Pl&R) (71152 - 1 Sample)

.1 Review of ltems Entered into the CAP

a. Inspection Scope

As required by lP 71152, "ldentification and Resolution of Problems," and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed screening of all items entered into the licensee's CAP. This was accomplished by reviewing the description of each new action request (AR) / lR and attending daily management review committee meetings.

Findinqs No findings were identified.

.2 Corrective Actions for Multiple Slow CRs (1 Annual Sample)

Inspection Scope ln January 2010, Exelon identified a total of 21 slow CRs at Unit 2 while performing scram time testing to meet TS surveillance requirements (SRs) 3.1

.4.2 and 3.1 .4.3 (see

NRC lnspection Report 0500027712010002, Section 1 R12). Exelon determined that the degradation of Viton-A scram solenoid pilot valve (SSPV) diaphragms, installed in 1995, caused the CRs to be slow between CR notch positions 48 and 46. This inspection focused on Exelon's problem identification, evaluation, and resolution associated with the Viton-A SSPV replacements at both units and scram time testing performance trending (Exelon lR 1023827).

The inspectors reviewed Exelon's associated root cause analysis (RCA), EOC review, and short and longterm corrective actions. The inspectors observed portions of SSPV replacement preventive maintenance on four, Unit 3 hydraulic control units (HCUs) and conducted several walkdowns of HCUs at both units to assess the material condition, maintenance practices, and configuration control. The inspectors also reviewed scram time testing results, performance monitoring and surveillance procedures, engineering evaluations, laboratory analysis reports, related industry operating experience (OE), and HCU maintenance history. The inspectors also directly inspected the internal diaphragms from three, Unit 3 SSPVs, installed in 1997 and removed in March 2011 , to independently validate Exelon's RCA assumptions associated with which SSPVs contained Viton-A diaphragms.

The inspectors reviewed a sample of CR related problems that Exelon identified and entered into the CAP since February 2010. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions. In addition, the inspectors reviewed corrective action lRs written on issues identified during the inspection to verify adequate problem identification and incorporation of the problem into the CAP. Documents reviewed are listed in the

.

b. Findinos and Observations No findings were identified. The inspectors concluded that Exelon had taken timely and appropriate action in accordance with the PBAPS TSs, operating and administrative procedures, and Exelon's CAP. The inspectors determined that Exelon's associated RCA was sufficiently thorough and based on testing (including independent laboratory analysis), sound engineering judgment, and relevant industry OE. Exelon's assigned corrective actions were aligned with the identified causal factors, adequately tracked, appropriately documented, and completed as scheduled.

Based on the documents reviewed, plant walkdowns, and discussions with engineering personnel, the inspectors noted that Exelon personnel identified problems and entered them into the CAP at a low threshold. In response to several minor issues identified by the inspectors, Exelon personnel promptly initiated lRs and/or took immediate action to address the issue, 40A3 Follow-up of Events and Notices of Enforcement Discretion (71153 - 3 Samples)

-l (Closed) LER 05000277/2010002-00, lmproperly Fastened Rod Hanger Results in Inoperable Subsystem of ESW ===

On September 22, 2010, Engineering personnel determined that a rod hanger (33H8-5143) supporting the discharge pipe of the 'A' ESW pump had not been carrying adequate pipe load prior to recent upgrades of the ESW discharge pipe support system during the week of September 13, 2010. lt was determined that the rod hanger condition, prior to September 13,2010, would have been unacceptable due to its degraded seismic capability. The degraded condition of rod hanger 33HB-S143 only affected the 'A' ESW subsystem during postulated seismic conditions. Since this condition was caused by inadequate original construction installation, this event is considered to be a condition prohibited by TSs due to one subsystem of ESW being inoperable for a time period longer than allowed by TS. The enforcement aspects of this LER review are documented in Section 4OA7. This LER is closed'

.2 (Ctosed) LER 05000277/2010003-00, Laboratory Analysis ldentifies Safety Relief Valve

(SRV) and Safety Valve (SV) Set Point Deficiencies (1 Sample)

Based on information received between September 27 and September 30,2010, from a laboratory performing SRV and SV as-found testing, PBAPS personnel determined that SRV and SV setpoint deficiencies existed with two SRVs and one SV that were installed during the 18th operating cycle for Unit 2. The two SRVs and one SV were determined to have their as-found setpoints in excess of the TS allowable + 1 percent tolerance.

The two SRVs and one SV outside of their TS allowable range were within the American Society of Mechanical Engineers (ASME) Code allowable of + 3 percent tolerance. The cause of these valves being outside of their allowable as-found setpoints is due to setpoint drift. The SRVs and SVs were replaced with refurbished valves for the 19th Unit 2 operating cycle. Additionally, LER 2-10-3 stated that PBAPS will pursue a change to the plant's licensing bases to increase SRV and SV setpoint tolerances. The licensee documented the event in lR 1120516. There were no actual safety consequences associated with this event. This LER reported three previous LERs (3-07-01 ,2-06'02, and 3-05-04) that involved a total of eight SRVs and SVs exceeding their TS + 1 percent set point requirement due to setpoint drift. LER 3-07-01 stated that to be more consistent with industry practices, changes to the PBAPS licensing basis would be considered to allow for SRV and SV setpoint tolerances of + 3 percent as allowed by the ASME code. The enforcement aspects of this LER review are documented in Section 4OA7. This LER is closed.

.3 Event Notice #46373: HPCI Declared Inoperable (1 Sample)

a. Inspection Scope

On March 16, 2011, PBAPS personnel informed the inspectors that an event notification report was planned to meet the requirements of 10 CFR 50.72(bX3)(vXD).

Subsequently, on March 17,2011, Event Notice #46677, reported that PBAPS Unit 2 declared the HPCI system inoperable for a condition found during testing which could cause the system to malfunction when swapping suction sources. While Unit 2 HPCI was aligned to the suppression pool suction flow path, unsatisfactory results were obtained while venting for system fill verification, indicating potential voiding of a portion of the pump discharge piping. Unit 2 HPCI remained available while aligned to its normal suction, the condensate storage tank.

At the end of the inspection period, PBAPS's evaluation of this event, under lR 1 188457, was ongoing.

b. Findinqs No findings were identified.

4OA5 Other Activities

.1 (Closed) Unresolved lssue (URl) 05000277. 27812010004-02, Potentially Inadequate FH

Procedures Lead to Personnel Performance Errors While Handling Fuel

a. Inspection Scope

URI 05000277 , 27812010004-02 documented the potential procedure inadequacy issues that allowed inadequate coordination of simultaneous close proximity activities within the reactor vessel during core alterations and personnel performance error issues while handling fuel in the reactor core and the SFP. These events appeared to be examples where inadequate procedures contributed to FH issues. The issue was unresolved pending completion of PBAPS's investigation and cause evaluation processes under the CAP. The inspectors also reviewed corrective action documents (lRs 1 115Q41, 1117854,1114828, and 1 117251) that are listed in detail in the Attachment to this report.

In addition, the inspectors discussed the identified problems and evaluation activities with PBAPS personnel.

b. Findinos lntroduction: A Green self-revealing NCV of TS 5.4.1 "Procedures" was identified, because PBAPS's procedures for refueling equipment operation and core alterations were inadequate to prevent a fuel bundle from contacting a CSI submarine device, while the fuel bundle was being transported from the core to the SFP. In particular, SO procedure 18.1.A-2, "Operation of Refueling Platform," and FH procedure 6C, "Core Component - Core Transfers," Revision 63, did not provide sufficient procedure steps, precautions or human performance tools to prevent contact while the refueling platform was operated in the automatic mode and when core components were in close proximity to obstructions and interferences.

Description:

On September 19,2010, during the performance of fuel movement number 302 of P2R18 core shuffle 1, a fuel bundle (JLM491) contacted a CSI submarine device while being transported from the core to the SFP. The refueling crew initiated the move with the refueling platform in the manual mode due to the close proximity of the fuel bundle location to the CSI submarine. The close proximity of the fuel moves to the CSI submarine was due to a change in the fuel movement methodology that allowed fuel moves to occur across all four quadrants of the core during P2R18, instead of moving fuel on a quadrant by quadrant basis which had been the normal practice for previous outages.

The fuel bundle and refueling platform mast were directly adjacent to the CSI submarine's umbilical cord as the fuel bundle was hoisted out of the its core location.

Once the refueling platform operator (RPO) believed he was clear of the CSI submarine, he changed the operating mode of the refueling platform from manual to automatic.

Then, the refueling platform proceeded in the automatic mode towards the SFP and the fuel bundle made contact with a thruster attached to the rear of the CSI submarine.

On September 18, 2010, during fuel moves for P2R18 core shuffle 1, a second safety spotter on the bridge had the bridge stopped to avoid making contact with the CSI submarine. The refueling platform was in automatic operation at the time of this near-miss event, and it was noted in lR 1 114828 that the crew's failure to anticipate the path of the bridge in automatic may have contributed to this near-miss condition occurring.

The crew stopped fuel movement to review the event. The crew reevaluated whether it was more appropriate to use manual control while in the vicinity of the CSI submarine.

After a review of the fuel and core component handling procedures, the inspectors noted that the procedures did not require a dedicated safety spotter for refueling bridge operations in close proximity to the CSI submarine or other in-vessel obstructions and interferences not protected by the boundary zone computer. Also, the procedures did not provide the refueling platform crew with guidance regarding when manual operation of the platform would be required in lieu of automatic operation or whether independent verification or supervisor approval would be required for changing the refueling platform's mode of operation. The inspectors concluded that the corrective actions for this near-miss event were inadequate to prevent the collision event that occurred on the following night, September 19, 2010.

PBAPS performed a causal evaluation of the September 19,2010, event that documented that the refueling crew was aware of the proximity of the fuel moves to the CSI submarine. lt was noted that the crew decided to place the refueling platform in the manual mode; however, they did not discuss when the refueling platform could be placed back into the automatic mode and no crew member was assigned to verify a clear path from the core to the SFP. The evaluation also noted that, prior to returning to their stations to check the next location for fuel bundle JLM491, the LSRO and fuel spotter did not assist the RPO in verifying that the travel path was clear. Due to the close proximity of the fuel bundle to the CSI submarine, a fourth crew member, a safety spotter, did not have adequate time to warn the RPO of the impending contact.

The inspectors reviewed SO procedure 18.1.A-2, "Operation of Refueling Platform," and noted an inconsistency in the requirements for assigning safety spotters. Step 2.7 requires a dedicated safety spotter when the reactor cavity work platform (RCWP) hoist is installed inside the refueling platform boundary zone, but Step 3.14 only requires consideration of a dedicated safety spotter if unique equipment, such as in-vessel inspection or repair equipment may become an obstruction. Step 3.10.2 requires travel paths to be clear before moving core components, but neither SO 18.1 .A-2 nor FH procedure 6C, "Core Component - Core Transfers," specify the crew member(s)responsible for this step. The procedures also do not specify whether human performance tools, such as, peer, independent, or concurrent verification of a clear path is required before refueling plafform movement is commenced.

fhe inspectors also noted that SO 18.1.A-2 did not provide guidance or requirements regarding circumstances when the manual and automatic modes of refueling platform movement should used. A note associated with Step 4.7.11 states that the platform should be placed in automatic immediately following verification of grapple engagement for maximum efficiency of refueling platform movement. Although the associated caution states that initiating automatic operation prior to verifying a clear path may result in core component contact, as the above noted, the procedure does not assign responsibility for determining or verifying a clear path. Based on the above, the inspectors concluded that PBAPS's FH procedures, as implemented, did not provide sufficient barriers or defense-in-depth to prevent the fuel bundle from contacting the CSI submarine device.

Analysis:

The performance deficiency was that PBAPS's procedures for refueling equipment operation and core alterations were inadequate to prevent a fuel bundle from making contact with a CSI submarine device, while the fuel bundle was being transported from the core to the SFP. The inspectors determined that the finding was more than minor because the finding was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone's objective to provide reasonable assurance that physical design barriers (i.e., fuel cladding) protect the public from radionuclide releases caused by accidents or events. Although no fuel damage occurred during this event, the inadequate procedure resulted in a FH event that could have impacted the cladding and affected the cornerstone's objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. IMC 0609, "SDP," Attachment 0609.04, "Phase 1 - Initiat Screening and Characterization of Findings," was used to evaluate the significance of the finding. Attachment 0609.04, Table 4a, was used to evaluate the impact of the finding on fuel clad integrity. Appendix G was considered for the evaluation, but was not used because it does not directly address fuel clad integrity.

Based on the results of fuel sipping done in February 2011, PBAPS concluded that there was no damage to the clad integrity of the impacted fuel bundle that was permanently discharged to the SFP. Since the finding did not affect SFP cooling or inventory and since there was no damage to fuel clad integrity from the impact with the CSI submarine, the finding was determined to be of very low safety significance (Green).

The finding has a cross-cutting aspect in Human Error Prevention Techniques in the Work Practices component of the Human Performance area. Specifically, PBAPS FH procedures did not require human error prevention techniques that were commensurate with the risk of moving fuel in close proximity to obstructions and interferences (H.a(a)).

Enforcement:

TS 5.4.1, "Procedures," requires that written procedures shall be established, implemented, and maintained covering the activities recommended in NRC Regulatory Guide (RG) 1.33, "Quality Assurance (aA) Program Requirements,"

Appendix A, November 1972. RG 1.33, Appendix A, Section B, "General Plant Operating Procedures," specifies procedures for Refueling Equipment Operation and Core Alterations. Contrary to the above, on September 19, 2010, during performance of move 3Q2 of P2R18 core shuffle 1, PBAPS's procedures for refueling equipment operation and core alterations, as established, implemented and maintained, were inadequate to prevent a fuel bundle from contacting a CSI submarine device, while the fuel bundle was being transported from the core to the SFP. Specifically, SO procedure 18.1.A-2, "Operation of Refueling Platform," Revision 22, and FH procedure 6C, "Core Component - Core Transfers," Revision 63, did not provide sufficient procedure steps, precautions or human performance tools necessary to ensure that fuel and core components would not encounter any obstructions or interferences. In particular, the procedures, as implemented, were inadequate to prevent contact while the refueling platform was operated in the automatic mode and when core components were in close proximity to obstructions and interferences not protected by the boundary zone computer. Corrective actions included entering the issue into the CAP, stopping FH until a prompt investigation was completed, and briefing crews on the event before FH resumed. Because this finding was of very low safety significance (Green) and PBAPS has entered it into their CAP via lR 1115041, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000277, 278t2011002-01, FH Procedures Were Inadequate to Prevent Fuelfrom Gontacting an Obstruction)

.2 Cask #50 Lid Seal Weld Repair (60853 - 1 Sample)

Backqround Cask #50 was loaded with spent fuel in accordance with Certificate of Compliance (CoC)1027, Amendment 1, issued October 30, 2007. The loaded cask was placed on the ISFSI pad in May 2010. On September 4,2010, an alarm was received indicating low helium pressure for Cask #50. On September 5,2010, the helium over pressure system was measured to be 40 pounds per square inch gauge (psig) and was then recharged to 75 psig. On September 9, 2010, the cask was transported to the refuel floor of Unit 3.

The licensee put in place a monitoring plan to record and chart the helium pressure of the cask on a daily basis. The monitoring program revealed the cask continued to slowly leak helium. The licensee began troubleshooting and on October 22,2010, the licensee was able to identify the source of the helium leak as the lid seal weld. The lid seal weld is performed during the cask manufacturing stage, therefore, it is considered a manufacturing defect. The licensee worked with the cask manufacturer, Transnuclear, lnc., to prepare a repair plan.

a. lnspection Scope The inspectors were on-site February 3-4, 2011, to perform an inspection of the repair of the lid seal weld. The inspectors observed and evaluated the welding and nondestructive examination (NDE) to determine whether the Peach Bottom staff and contractors had developed the capability to properly repair and perform NDE of a pin hole size leak identified in a seal weld of a one-half inch diameter plug in the lid of ISFSI cask TN-68-50-A. Because the plug being repaired was similar to two other seal-welded plugs, the inspectors reviewed the work to confirm adequacy of these other two plugs.

The inspectors observed the process for locating the leaking plug, grinding for leak removal, the welding equipment setup, welding of the plug repair area, the magnetic particle testing (MT) equipment, and the helium leak testing station. The scope, plans, and equipment setup for helium leak testing of the plug areas, and the helium leak testing were reviewed. The materials used for welding were verified to be compatible with the lid and plug compositions, and were confirmed to meet the welding procedure.

The inspectors also examined the welding equipment, observed welding in progress and observed the weld surface. The inspectors reviewed the controls on localized temperature increase by preheating and welding and noted that limits to the extent of heating were established by a specific engineering-based calculation. The inspectors discussed the work steps and plans with those involved in the repair of the weld. The inspection included verification that the activities were accomplished in accordance with the commitments and requirements contained in the Safety Analysis Report (SAR), the NRC's SER, the CoC, the ASME Code, as well as the licensee's QA program, and 10 CFR Part 72.

The inspectors reviewed the repair plan, 10 CFR 72.48 review, cask lid drawings, welder qualification records, procedures for welding, visual weld examination, MT testing, loading and transport operations, and helium leak tests. The calculation for the thermal analysis of the seal weld rework was reviewed by an engineer in the Office of Nuclear Material Safety and Safeguards (NMSS).

b. Findinos No findings were identified.

40A6 Meetinqs. Includinq Exit

.1 Quarterlv Resident Exit Meetinq Summarv

On April 29,2Q11, the resident inspectors presented the inspection results to Mr. Thomas Dougherty and other PBAPS staff, who acknowledged the findings.

Mr, P. Krohn, Chief, USNRC, Region 1, Division of Reactor Projects, Branch 4, attended this quarterly inspection exit meeting. The inspectors asked the licensee whether any of the information discussed as being included in the report should be considered proprietary. No proprietary information was identified.

.2 Manaqement Meetinqs

The inspection results for the inspection of the Spent Fuel Cask #50 lid seal weld repair were discussed with Garey Stathes, Plant Manager, and other members of the PBAPS staff via teleconference on March 15,2011.

The inspection results for the inspection of the Licensed Operator Requalification Program were discussed on March 11,2011, with members of the PBAPS staff.

The inspection results for the inspection of the Peach Bottom EP Pls were discussed on February 18,2011, with Mr. T. Dougherty, Site Vice President, and other members of the PBAPS staff. After the inspectors conducted the Exelon Mid-Atlantic corporate EP inspection, an exit meeting was conducted on February 22,2011, with V. Cwietniewicz, Mid-Atlantic EP Manager, and other plant staff to discuss the results and observations of the corporate inspection.

The inspection results for the inspection of the PBAPS SSPV issues were discussed on March 25,2011, with Mr. Thomas Dougherty, Site Vice President, and other members of Exelon management. The inspector verified that no proprietary information is documented in this feeder.

4AA7 Licensee-ldentified Violations The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as a NCV:

ln Modes 1,2 and 3, with one ESW subsystem inoperable for more than seven days, TS Limiting Condition for Operation (LCO) 3.7.2, condition C, requires the unit to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Contrary to the above, since original construction and prior to September 13,2010, an engineering evaluation determined that the 'A' ESW subsystem was inoperable due to the degraded seismic capability of rod hanger 33HB-S143 that only affected the'A'ESW subsystem. During upgrades to the ESW discharge pipe support system during the week of September 13, 201Q, PBAPS personnel identified that the original installation of the rod hanger had not been carrying adequate pipe load. This condition was considered as a condition prohibited by TS due to one subsystem of ESW being inoperable for greater than the time period allowed by TS. The cause of the event was due to an inadequate design drawing. PBAPS documented this issue in the CAP as lRs 1114812 and 1118711. Since there was no actual loss of safety function as a result of this event, this issue is of very low (Green) safety significance. The LER associated with the event was documented in Section 4OA3.1.

TS LCO 3.4.3 requires the safety function of 11 valves (any combination of SRVs and SVs) to be operable during operational Modes 1, 2, and 3 or else be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within an additional 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Contrary to the above, two SRVs and one SV were determined to have their as-found setpoints in excess of the TS allowable tolerance, thus leaving 10 operable SRVs and SVs.

The SRVs and SVs were replaced with refurbished valves for the 19th Unit 2 operating cycle. Additionally, LER 2-10-3 stated that PBAPS will pursue a change to the plant's licensing bases to increase SRV and SV setpoint tolerances to the ASME Code allowable + 3 percent tolerance. The licensee documented the event in lR 1 120516. Since there was no actual loss of safety function as a result of this event, this issue is of very low (Green) safety significance. The LER associated with the event was documented in Section 4c.A3.2.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Exelon Generation Comoanv Personnel

T. Dougherty, Site Vice President

G. Stathes, Plant Manager

J. Armstrong, Regulatory Assurance Manager

T. Moore, Site Engineering Director

P. Navin, Operations Director

J. Kovalchick, Security Manager

P. Cowan, Work Management Director

L. Lucas, Chemistry Manager

R. Holmes, Radiation Protection Manager

T. Wasong, Training Director

C. Goff, Operations Training Manager

NRC Personnel

P. Krohn, Branch Chief

F. Bower, Senior Resident Inspector

A. Ziedonis, Resident Inspector

C. Crisden, Emergency Preparedness Specialist

J. D'Antonio, Senior Operations Engineer

T. Fish, Senior Operations Engineer

S. Hammann, Senior Health Physicist

J. Schoppy, Senior Reactor Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Opened/Closed

05000277, 2781201 1002-01 NCV FH Procedures Were Inadequate to Prevent Fuelfrom Contacting an Obstruction (Section 4OA5. 1 )

Closed

0500027712010002-00 LER lmproperly Fastened Rod Hanger Results in Inoperable Subsystem of ESW (Section 4OA3.1)

05000277 t2010003-00 LER Laboratory Analysis ldentifies SRV and SV Set Point Deficiencies (Section 4C.43.2)

05000277, 27 812010004-02 uRl Potentially Inadequate FH Procedures Lead to Personnel Performance Errors While Handling Fuel (Section 4OA5.1)

Discussed

None